Technical assessment of continued wet storage of EBR-II fuel

International Nuclear Information System (INIS)

Pahl, R.G.; Franklin, E.M.; Ebner, M.A.

1996-01-01

A technical assessment of the continued wet storage of EBR-II fuel has been made. Previous experience has shown that in-basin cladding failure occurs by intergranular attack of sensitized cladding, likely assisted by basin water chlorides. Subsequent fuel oxidation is rapid and leads to loss of configuration and release of fission products. The current inventory of EBR-II fuel stored in the ICPP basins is at risk from similar corrosion reactions

In-reactor cladding breach of EBR-II driver-fuel elements

International Nuclear Information System (INIS)

Seidel, B.R.; Einziger, R.E.

1977-01-01

Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

Alternate-Fueled Combustor-Sector Performance: Part A: Combustor Performance Part B: Combustor Emissions

Science.gov (United States)

Shouse, D. T.; Neuroth, C.; Henricks, R. C.; Lynch, A.; Frayne, C.; Stutrud, J. S.; Corporan, E.; Hankins, T.

2010-01-01

Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as drop-in fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of feedstock. Adherence to alternate fuels and fuel blends requires smart fueling systems or advanced fuel-flexible systems, including combustors and engines without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data for synthetic-parafinic-kerosene- (SPK-) type fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling.

Fuel cycle math - part two

International Nuclear Information System (INIS)

Anon.

1992-01-01

This article is Part 2 of a two part series on simple mathematics associated with the nuclear fuel cycle. While not addressing any of the financial aspects of the fuel cycle, this article does discuss the following: conversion between English and metric systems; uranium content expressed in equivalent forms, such as U3O8, and the method of determining these equivalencies; the uranium conversion process, considering different input and output compounds; and the enrichment process, including feed, tails, and product assays, as well as SWU and feed requirements

Off-normal performance of EBR-II [Experimental Breeder Reactor] driver fuel

International Nuclear Information System (INIS)

Seidel, B.R.; Batte, G.L.; Lahm, C.E.; Fryer, R.M.; Koenig, J.F.; Hofman, G.L.

1986-09-01

The off-normal performance of EBR-II Mark-II driver fuel has been more than satisfactory as demonstrated by robust reliability under repeated transient overpower and undercooled loss-of-flow tests, by benign run-beyond-cladding-breach behavior, and by forgiving response to fabrication defects including lack of bond. Test results have verified that the metallic driver fuel is very tolerant of off-normal events. This behavior has allowed EBR-II to operate in a combined steady-state and transient mode to provide test capability without limitation from the metallic driver fuel

IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

International Nuclear Information System (INIS)

Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

1991-01-01

The next major milestone of the IFR (Integral Fast Reactor) program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. (author)

IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

International Nuclear Information System (INIS)

Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

1991-01-01

The next major milestone of the IFR program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. 5 refs., 4 figs

Carbonate-mediated Fe(II) oxidation in the air-cathode fuel cell: a kinetic model in terms of Fe(II) speciation.

Science.gov (United States)

Song, Wei; Zhai, Lin-Feng; Cui, Yu-Zhi; Sun, Min; Jiang, Yuan

2013-06-06

Due to the high redox activity of Fe(II) and its abundance in natural waters, the electro-oxidation of Fe(II) can be found in many air-cathode fuel cell systems, such as acid mine drainage fuel cells and sediment microbial fuel cells. To deeply understand these iron-related systems, it is essential to elucidate the kinetics and mechanisms involved in the electro-oxidation of Fe(II). This work aims to develop a kinetic model that adequately describes the electro-oxidation process of Fe(II) in air-cathode fuel cells. The speciation of Fe(II) is incorporated into the model, and contributions of individual Fe(II) species to the overall Fe(II) oxidation rate are quantitatively evaluated. The results show that the kinetic model can accurately predict the electro-oxidation rate of Fe(II) in air-cathode fuel cells. FeCO3, Fe(OH)2, and Fe(CO3)2(2-) are the most important species determining the electro-oxidation kinetics of Fe(II). The Fe(II) oxidation rate is primarily controlled by the oxidation of FeCO3 species at low pH, whereas at high pH Fe(OH)2 and Fe(CO3)2(2-) are the dominant species. Solution pH, carbonate concentration, and solution salinity are able to influence the electro-oxidation kinetics of Fe(II) through changing both distribution and kinetic activity of Fe(II) species.

Fuel cells (part 2)

International Nuclear Information System (INIS)

Campanari, S.; Macchi, E.

1999-01-01

The article, following and completing the issues dealt with in part 1 (CH4 Energia Metano, 1/99, p. 7), describe the operating characteristic and construction features of molten carbonate and solid oxide fuel cells (MCFC and SOFC). For the latter type, construction cost are evaluated by various authors and research institutes. The article ends by presenting some tables showing the classification and the main characteristics of various fuel cells, and well as the effect of some gases on the behaviour of some of them [it

Fuel Quality/Processing Study. Volume II. Appendix, Task I, literature survey

Energy Technology Data Exchange (ETDEWEB)

O' Hara, J B; Bela, A; Jentz, N E; Klumpe, H W; Kessler, R E; Kotzot, H T; Loran, B I

1981-04-01

This activity was begun with the assembly of information from Parsons' files and from contacts in the development and commercial fields. A further more extensive literature search was carried out using the Energy Data Base and the American Petroleum Institute Data Base. These are part of the DOE/RECON system. Approximately 6000 references and abstracts were obtained from the EDB search. These were reviewed and the especially pertinent documents, approximately 300, were acquired in the form of paper copy or microfiche. A Fuel Properties form was developed for listing information pertinent to gas turbine liquid fuel properties specifications. Fuel properties data for liquid fuels from selected synfuel processes, deemed to be successful candidates for near future commercial plants were tabulated on the forms. The processes selected consisted of H-Coal, SRC-II and Exxon Donor Solvent (EDS) coal liquefaction processes plus Paraho and Tosco shale oil processes. Fuel properties analyses for crude and distillate syncrude process products are contained in Section 2. Analyses representing synthetic fuels given refinery treatments, mostly bench scale hydrotreating, are contained in Section 3. Section 4 discusses gas turbine fuel specifications based on petroleum source fuels as developed by the major gas turbine manufacturers. Section 5 presents the on-site gas turbine fuel treatments applicable to petroleum base fuels impurities content in order to prevent adverse contaminant effects. Section 7 relates the environmental aspects of gas turbine fuel usage and combustion performance. It appears that the near future stationary industrial gas turbine fuel market will require that some of the synthetic fuels be refined to the point that they resemble petroleum based fuels.

Fuel cycle math - part one

International Nuclear Information System (INIS)

Anon.

1992-01-01

This article is Part One of a two-part article that reviews some of the numbers associated with the nuclear fuel cycle. The contents of Part One include: composition of the element uranium, considering atomic mass and weight-percent of the isotopes; uranium in the ground, including ore grades; mining, with dilution factors and recovery rates; ore sorting, including concentration factors; and uranium recovery. No financial information is presented in either Part One or Part Two

Compatibility analysis of DUPIC fuel (Part II) - Reactor physics design and analysis

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Choi, Hang Bok; Rhee, Bo Wook; Roh, Gyu Hong; Kim, Do Hun [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

The compatibility analysis of the DUPIC fuel in a CANDU reactor has been assessed. This study includes the fuel composition adjustment, comparison of lattice properties, performance analysis of reactivity devices, determination of regional over-power (ROP) trip setpoint, and uncertainty estimation of core performance parameters. For the DUPIC fuel composition adjustment, three options have been proposed, which can produce uniform neutronic characteristics of the DUPIC fuel. The lattice analysis has shown that the characteristics of the DUPIC fuel is compatible with those of natural uranium fuel. The reactivity devices of the CANDU-6 reactor maintain their functional requirements even for the DUPIC fuel system. The ROP analysis has shown that the trip setpoint is not sacrificed for the DUPIC fuel system owing to the power shape that enhances more thermal margin. The uncertainty analysis of the core performance parameter has shown that the uncertainty associated with the fuel composition variation is reduced appreciably, which is primarily due to the fuel composition adjustment and secondly the on-power refueling feature and spatial control function of the CANDU reactor. The reactor physics calculation has also shown that it is feasible to use spent PWR fuel directly in CANDU reactors without deteriorating the CANDU-6 core physics design requirements. 29 refs., 67 figs., 60 tabs. (Author)

EBR-II fuel handling console digital upgrade

International Nuclear Information System (INIS)

Peters, G.G.; Wiege, D.D.; Christensen, L.J.

1995-01-01

The main fuel handling console and control system at the Experimental Breeder Reactor II (EBR-II) are being upgraded to a computerized system using high-end workstations for the operator interface and a programmable logic controller (PLC) for the control system. Two-dimensional (2D) and three-dimensional (3D) computer graphics will be provided for the operator which will show the relative position of under-sodium fuel handling equipment. This equipment is operated remotely with no means of directly viewing the transfer. This paper describes various aspects of the modification including reasons for the upgrade, capabilities the new system provides over the old control system, philosophies and rationale behind the new design, testing and simulation work, diagnostic features, and the advanced graphics techniques used to display information to the operator

Performance of advanced oxide fuel pins in EBR-II

International Nuclear Information System (INIS)

Lawrence, L.A.; Jensen, S.M.; Hales, J.W.; Karnesky, R.A.; Makenas, B.J.

1986-05-01

The effects of design and operating parameters on mixed-oxide fuel pin irradiation performance were established for the Hanford Engineering Development Laboratory (HEDL) advanced oxide EBR-II test series. Fourteen fuel pins breached in-reactor with reference 316 SS cladding. Seven of the breaches are attributed to FCMI. Of the remaining seven breached pins, three are attributed to local cladding over-temperatures similar to the breach mechanism for the reference oxide pins irradiated in EBR-II. FCCI was found to be a contributing factor in two high burnup, i.e., 11.7 at. % breaches. The remaining two breaches were attributed to mechanical interaction of UO 2 fuel and fission products accumulated in the lower cladding insulator gap, and a loss of cladding ductility possibly due to liquid metal embrittlement. Fuel smear density appears to have the most significant impact on lifetime. Quantitative evaluations of cladding diameter increases attributed to FCMI, established fuel smear density, burnup, and cladding thickness-to-diameter ratio as the major parameters influencing the extent of cladding strain

Results of Cesar II critical facility with low enriched fuel balls

Energy Technology Data Exchange (ETDEWEB)

Langlet, G; Guerange, J; Laponche, B; Morier, F; Neef, R D; Bock, H J; Kring, F J; Scherer, W

1972-06-15

The Cesar facility has been transformed to load in its center a pebble bed fuel. This new Cesar assembly is called Cesar II. The program for the measurements with HTR type fuel balls is managed under a cooperation between physicists of CEA/CADARACHE and KFA/JUELICH. A description of the measuring zones of Cesar II and of the experimental results is given.

Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel

International Nuclear Information System (INIS)

Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

2010-01-01

A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

International Nuclear Information System (INIS)

Hilton, B.A.

2000-01-01

The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

Safety aspects of advanced fuels irradiations in EBR-II

International Nuclear Information System (INIS)

Lehto, W.K.

1975-09-01

Basic safety questions such as MFCI, loss-of-Na bond, pin behavior during design basis transients, and failure propagation were evaluated as they pertain to advanced fuels in EBR-II. With the exception of pin response to the unlikely loss-of-flow transient, the study indicates that irradiation of significant numbers of advanced fueled subassemblies in EBR-II should pose no safety problems. The analysis predicts, however, that Na boiling may occur during the postulated design basis unlikely loss-of-flow transient in subassemblies containing He-bonded fuel pins with the larger fuel-clad gaps. The calculations indicate that coolant temperatures at top of core in the limiting S/A's, containing the He bonded pins, would reach approximately 1480 0 F during the transient without application of uncertainty factors. Inclusion of uncertainties could result in temperature predictions which approach coolant boiling temperatures (1640 0 F). Further analysis of He-bonded pins is being done in this potential problem area, e.g., to apply best estimates of uncertainty factors and to determine the sensitivity of the preliminary results to gap conductance

System modeling of spent fuel transfers at EBR-II

International Nuclear Information System (INIS)

Imel, G.R.; Houshyar, A.

1994-01-01

The unloading of spent fuel from the Experimental Breeder Reactor-II (EBR-II) for interim storage and subsequent processing in the Fuel Cycle Facility (FCF) is a multi-stage process, involving complex operations at a minimum of four different facilities at the Argonne National Laboratory-West (ANL-W) site. Each stage typically has complicated handling and/or cooling equipment that must be periodically maintained, leading to both planned and unplanned downtime. A program was initiated in October, 1993 to replace the 330 depleted uranium blanket subassemblies (S/As) with stainless steel reflectors. Routine operation of the reactor for fuels performance and materials testing occurred simultaneously in FY 1994 with the blanket unloading. In the summer of 1994, Congress dictated the October 1, 1994 shutdown of EBR-2. Consequently, all blanket S/As and fueled drivers will be removed from the reactor tank and replaced with stainless steel assemblies (which are needed to maintain a precise configuration within the grid so that the under sodium fuel handling equipment can function). A system modeling effort was conducted to determine the means to achieve the objective for the blanket and fuel unloading program, which under the current plan requires complete unloading of the primary tank of all fueled assemblies in 2 1/2 years. A simulation model of the fuel handling system at ANL-W was developed and used to analyze different unloading scenarios; the model has provided valuable information about required resources and modifications to equipment and procedures. This paper reports the results of this modeling effort

Fuel behavior in advanced water reactors

International Nuclear Information System (INIS)

Bolme, A.B.

1996-01-01

Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel; Metodo Optimizado de Extraccion por Ultrasonidos para la Determinacion de PAHs Seleccionados en un Suelo Contaminado con Fuel de Tipo II

Energy Technology Data Exchange (ETDEWEB)

Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

2010-10-21

A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

Performance of commercially produced mixed-oxide fuels in EBR-II

International Nuclear Information System (INIS)

Hales, J.W.; Lawrence, L.A.

1980-11-01

Commercially produced fuels for the Fast Flux Test Facility (FFTF) were irradiated in EBR-II under conditions of high cladding temperature (approx. 700 0 C) and low power (approx. 200 W/cm) to verify that manufacturing processes did not introduce variables which significantly affect general fuel performance. Four interim examinations and a terminal examination were completed to a peak burnup of 5.2 at. % to provide irradiation data pertaining to fuel restructuring and dimensional stability at low fuel temperature, fuel-cladding reactions at high cladding temperature and general fuel behavior. The examinations indicate completely satisfactory irradiation performance for low heat rates and high cladding temperatures to 5.2 at. % burnup

Temperature feedback of TRIGA MARK-II fuel

Science.gov (United States)

Usang, M. D.; Minhat, M. S.; Rabir, M. H.; M. Rawi M., Z.

2016-01-01

We study the amount of temperature feedback on reactivity for the three types of TRIGA fuel i.. ST8, ST12 and LEU fuel, are used in the TRIGA MARK II reactor in Malaysia Nuclear Agency. We employ WIMSD-5B for the calculation of kin f for a single TRIGA fuel surrounded by water. Typical calculations of TRIGA fuel reactivity are usually limited to ST8 fuel, but in this paper our investigation extends to ST12 and LEU fuel. We look at the kin f of our model at various fuel temperatures and calculate the amount reactivity removed. In one instance, the water temperature is kept at room temperature of 300K to simulate sudden reactivity increase from startup. In another instance, we simulate the sudden temperature increase during normal operation where the water temperature is approximately 320K while observing the kin f at various fuel temperatures. For accidents, two cases are simulated. The first case is for water temperature at 370K and the other is without any water. We observe that the higher Uranium content fuel such as the ST12 and LEU have much smaller contribution to the reactivity in comparison to the often studied ST8 fuel. In fact the negative reactivity coefficient for LEU fuel at high temperature in water is only slightly larger to the negative reactivity coefficient for ST8 fuel in void. The performance of ST8 fuel in terms of negative reactivity coefficient is cut almost by half when it is in void. These results are essential in the safety evaluation of the reactor and should be carefully considered when choices of fuel for core reconfiguration are made.

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Results of tests with open fuel in KNK II

International Nuclear Information System (INIS)

Schmitz, G.

1987-03-01

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

The continual fuel management modification in Qinshan project II

International Nuclear Information System (INIS)

Ye Guodong; Pan Zefei; Zhang Xingtian

2010-01-01

The fuel management strategy is the basis of the nuclear power plants. The performance of the fuel management strategy affects the plants' safety and economy indicators directly. The paper summarizes all the modifications on the fuel management work in Qinshan Project II since the plant was established. It includes the surveillance system of physics tests, fetching in high performance fuel assemblies, reloading pattern optimization, and the modifications of the final safety analysis report. At the same time, it evaluates the benefit of the modifications in the few years. The experience in this paper is much helpful and could be implemented on the same type plants. (authors)

The EBR-II fuel cycle story

International Nuclear Information System (INIS)

Stevenson, C.E.

1987-01-01

This volume on the history of the Experimental Breeder Reactor (EBR) program and the Fuel Cycle Facility (FCF) offers both the historical perspective and ''reasons why'' the project was so successful. The operation of the FCF in conjunction with the EBR-II was prepared because of the unique nature of the pyrmetallurgical processing system that was demonstrated at the time. Following brief descriptions and histories of the EBR-I and EBR-II reactors, the FCF and its process requirements are described. The seven principal process steps are presented, including for each one, the development, equipment used, operating procedures, results, problems and other data. Scrap and waste disposition, analytical control, safety, management, and cost of the FCF are also included

Fuel cell-gas turbine hybrid system design part II: Dynamics and control

Science.gov (United States)

McLarty, Dustin; Brouwer, Jack; Samuelsen, Scott

2014-05-01

Fuel cell gas turbine hybrid systems have achieved ultra-high efficiency and ultra-low emissions at small scales, but have yet to demonstrate effective dynamic responsiveness or base-load cost savings. Fuel cell systems and hybrid prototypes have not utilized controls to address thermal cycling during load following operation, and have thus been relegated to the less valuable base-load and peak shaving power market. Additionally, pressurized hybrid topping cycles have exhibited increased stall/surge characteristics particularly during off-design operation. This paper evaluates additional control actuators with simple control methods capable of mitigating spatial temperature variation and stall/surge risk during load following operation of hybrid fuel cell systems. The novel use of detailed, spatially resolved, physical fuel cell and turbine models in an integrated system simulation enables the development and evaluation of these additional control methods. It is shown that the hybrid system can achieve greater dynamic response over a larger operating envelope than either individual sub-system; the fuel cell or gas turbine. Results indicate that a combined feed-forward, P-I and cascade control strategy is capable of handling moderate perturbations and achieving a 2:1 (MCFC) or 4:1 (SOFC) turndown ratio while retaining >65% fuel-to-electricity efficiency, while maintaining an acceptable stack temperature profile and stall/surge margin.

Metal waste forms from treatment of EBR-II spent fuel

International Nuclear Information System (INIS)

Abraham, D. P.

1998-01-01

Demonstration of Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel is currently being conducted on irradiated, metallic driver fuel and blanket fuel elements from the Experimental Breeder Reactor-II (EBR-II) in Idaho. The residual metallic material from the electrometallurgical treatment process is consolidated into an ingot, the metal waste form (MWF), by employing an induction furnace in a hot cell. Scanning electron microscopy (SEM) and chemical analyses have been performed on irradiated cladding hulls from the driver fuel, and on samples from the alloy ingots. This paper presents the microstructures of the radioactive ingots and compares them with observations on simulated waste forms prepared using non-irradiated material. These simulated waste forms have the baseline composition of stainless steel - 15 wt % zirconium (SS-15Zr). Additions of noble metal elements, which serve as surrogates for fission products, and actinides are made to that baseline composition. The partitioning of noble metal and actinide elements into alloy phases and the role of zirconium for incorporating these elements is discussed in this paper

Workshop 96. Part II

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-31

Part II of the seminar proceedings contains contributions in various areas of science and technology, among them materials science in mechanical engineering, materials science in electrical, chemical and civil engineering, and electronics, measuring and communication engineering. In those areas, 6 contributions have been selected for INIS. (P.A.).

Workshop 96. Part II

International Nuclear Information System (INIS)

1995-12-01

Part II of the seminar proceedings contains contributions in various areas of science and technology, among them materials science in mechanical engineering, materials science in electrical, chemical and civil engineering, and electronics, measuring and communication engineering. In those areas, 6 contributions have been selected for INIS. (P.A.)

Degradation of EBR-II driver fuel during wet storage

International Nuclear Information System (INIS)

Pahl, R. G.

2000-01-01

Characterization data are reported for sodium bonded EBR-II reactor fuel which had been stored underwater in containers since the 1981--1982 timeframe. Ten stainless steel storage containers, which had leaked water during storage due to improper sealing, were retrieved from the ICPP-603 storage basin at the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. In the container chosen for detailed destructive analysis, the stainless steel cladding on the uranium alloy fuel had ruptured and fuel oxide sludge filled the bottom of the container. Headspace gas sampling determined that greater than 99% hydrogen was present. Cesium 137, which had leached out of the fuel during the aqueous corrosion process, dominated the radionuclide source term of the water. The metallic sodium from the fuel element bond had reacted with the water, forming a concentrated caustic solution of NaOH

Prototypical spent nuclear nuclear fuel rod consolidation equipment, Phase 2: Final design report: Volume 2, Appendices: Part 1

International Nuclear Information System (INIS)

Ciez, A.P.

1987-01-01

The purpose of this specification is to establish functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation System. The Department of Energy-Idaho Operations Office (DOE-ID) is responsible for the implementation of the Prototypic Dry Rod Consolidation Demonstration Project. This program is to develop and demonstrate a fully qualified, licensable, cost-effective, dry spent fuel rod consolidation system by July 1989. The work is divided into four phases as follows: Phase I--Preliminary Design, Phase II--Final Design Option, Phase III--Fabrication and System Checkout Option, and Phase IV--Installation and Hot Demonstration Option. This specification is part of the Phase II effort. The objectives of this specification are to provide functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation equipment; establish specific tool and subsystem requirements such that the integrated and overall system requirements are satisfied; and establish positioning, envelope and size interface control requirements for each tool or subsystem such that the individual components will interface properly with the overall system design

Stationary liquid fuel fast reactor SLFFR — Part II: Safety analysis

Energy Technology Data Exchange (ETDEWEB)

Jing, T.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

2016-12-15

Highlights: • A multi-channel safety analysis code named MUSA is developed for SLFFR transient analyses. • MUSA is verified against the SYS4A/SASSYS-1 code by simulating the ULOF accident for the advanced burner test reactor. • It is shown that SLFFR has a passive shutdown capability for double-fault, beyond-design-basis accidents UTOP, ULOHS and ULOF. - Abstract: Safety characteristics have been evaluated for the stationary liquid fuel fast reactor (SLFFR) proposed for effective burning of hazardous TRU elements of used nuclear fuel. In order to model the geometrical configuration and reactivity feedback mechanisms unique to SLFFR, a multi-channel safety analysis code named MUSA was developed. MUSA solves the time-dependent coupled neutronics and thermal-fluidic problems. The thermal-fluidic behavior of the core is described by representing the core with one-dimensional parallel channels. The primary heat transport system is modeled by connecting compressible volumes by liquid segments. A point kinetics model with six delayed neutron groups is used to represent the fission power transients. The reactivity feedback is estimated by combining the temperature and density variations of liquid fuel, structural material and sodium coolant with the corresponding axial distributions of reactivity worth in each individual thermal-fluidic channel. Preliminary verification tests with a conventional solid fuel reactor agreed well with the reference solutions obtained with the SAS4A/SASSYS-1 code. Transient analyses of SLFFR were performed for unprotected transient over-power (UTOP), unprotected loss of heat sink (ULOHS) and unprotected loss of flow (ULOF) accidents. The results showed that the thermal expansion of liquid fuel provides sufficiently large negative feedback reactivity for passive shutdown of UTOP and ULOHS. The ULOF transient is also terminated passively with the negative reactivity introduced by the gas expansion modules installed at the core periphery

EBR-II: search for the lost subassembly

International Nuclear Information System (INIS)

King, R.W.; Buschman, H.W.; Poloncsik, J.; Remsburg, J.S.; Sine, H.W.

1983-01-01

Experimental Breeder Reactor II (EBR-II) has been operating for nearly 20 years as part of the foundation of the US Department of Energy's LMFBR development program. During that time, the EBR-II fuel-handling system has performed extremely well, especially considering the conditions under which much of the system operates and the reliability required to maintain the high plant factor routinely demonstrated by EBR-II. Since EBR-II is a pool-type reactor, much of the fuel handling is done remotely within the sodium-filled primary tank at 371 0 C. Activities involved in locating a misplaced fuel subassembly in the primary tank are described

Safety assessment for Dragon fuel element production

International Nuclear Information System (INIS)

Price, M.S.T.

1963-11-01

This report shall be the Safety Assessment covering the manufacture of the First Charge of Fuel and Fuel Elements for the Dragon Reactor Experiment. It is issued in two parts, of which Part I is descriptive and Part II gives the Hazards Analysis, the Operating Limitations, the Standing Orders and the Emergency Drill. (author)

Computer imaging of EBR-II fuel handling equipment

International Nuclear Information System (INIS)

Peters, G.G.; Hansen, L.H.

1995-01-01

This paper describes a three-dimensional graphics application used to visualize the positions of remotely operated fuel handling equipment in the EBR-II reactor. A three-dimensional (3D) visualization technique is necessary to simulate direct visual observation of the transfers of fuel and experiments into and out of the reactor because the fuel handling equipment is submerged in liquid sodium and therefore is not visible to the operator. The system described in this paper uses actual signals to drive a three-dimensional computer-generated model in real-time in response to movements of equipment in the plant This paper will present details on how the 3D model of the intank equipment was created and how real-time dynamic behavior was added to each of the moving components

Simulation of thermal stresses in anode-supported solid oxide fuel cell stacks. Part II: Loss of gas-tightness, electrical contact and thermal buckling

Science.gov (United States)

Nakajo, Arata; Wuillemin, Zacharie; Van herle, Jan; Favrat, Daniel

Structural stability issues in planar solid oxide fuel cells arise from the mismatch between the coefficients of thermal expansion of the components. The stress state at operating temperature is the superposition of several contributions, which differ depending on the component. First, the cells accumulate residual stresses due to the sintering phase during the manufacturing process. Further, the load applied during assembly of the stack to ensure electric contact and flatten the cells prevents a completely stress-free expansion of each component during the heat-up. Finally, thermal gradients cause additional stresses in operation. The temperature profile generated by a thermo-electrochemical model implemented in an equation-oriented process modelling tool (gPROMS) was imported into finite-element software (ABAQUS) to calculate the distribution of stress and contact pressure on all components of a standard solid oxide fuel cell repeat unit. The different layers of the cell in exception of the cathode, i.e. anode, electrolyte and compensating layer were considered in the analysis to account for the cell curvature. Both steady-state and dynamic simulations were performed, with an emphasis on the cycling of the electrical load. The study includes two different types of cell, operation under both thermal partial oxidation and internal steam-methane reforming and two different initial thicknesses of the air and fuel compressive sealing gaskets. The results generated by the models are presented in two papers: Part I focuses on cell cracking. In the present paper, Part II, the occurrences of loss of gas-tightness in the compressive gaskets and/or electrical contact in the gas diffusion layer were identified. In addition, the dependence on temperature of both coefficients of thermal expansion and Young's modulus of the metallic interconnect (MIC) were implemented in the finite-element model to compute the plastic deformation, while the possibilities of thermal buckling

49 CFR 536.10 - Treatment of dual-fuel and alternative fuel vehicles-consistency with 49 CFR part 538.

Science.gov (United States)

2010-10-01

... vehicles-consistency with 49 CFR part 538. 536.10 Section 536.10 Transportation Other Regulations Relating... vehicles—consistency with 49 CFR part 538. (a) Statutory alternative fuel and dual-fuel vehicle fuel... economy in a particular compliance category by more than the limits set forth in 49 U.S.C. 32906(a), the...

PWR Core II blanket fuel disposition recommendation of storage option study

International Nuclear Information System (INIS)

Dana, C.M.

1995-01-01

After review of the options available for current storage of T Plant Fuel the recommended option is wet storage without the use of chillers. A test has been completed that verifies the maximum temperature reached is below the industrial standard for storage of spent fuel. This option will be the least costly and still maintain the fuel in a safe environment. The options that were evaluated included dry storage with and without chillers, and wet storage with and without chillers. Due to the low decay heat of the Shippingport Core II Blanket fuel assemblies the fuel pool temperature will not exceed 100 deg. F

75 FR 29605 - Clean Alternative Fuel Vehicle and Engine Conversions

Science.gov (United States)

2010-05-26

... Part II Environmental Protection Agency 40 CFR Parts 85 and 86 Clean Alternative Fuel Vehicle and...-0299; FRL-9149-9] RIN 2060-AP64 Clean Alternative Fuel Vehicle and Engine Conversions AGENCY... streamline the process by which manufacturers of clean alternative fuel conversion systems may demonstrate...

EBR-II high-ramp transients under computer control

International Nuclear Information System (INIS)

Forrester, R.J.; Larson, H.A.; Christensen, L.J.; Booty, W.F.; Dean, E.M.

1983-01-01

During reactor run 122, EBR-II was subjected to 13 computer-controlled overpower transients at ramps of 4 MWt/s to qualify the facility and fuel for transient testing of LMFBR oxide fuels as part of the EBR-II operational-reliability-testing (ORT) program. A computer-controlled automatic control-rod drive system (ACRDS), designed by EBR-II personnel, permitted automatic control on demand power during the transients

Unlearning Established Organizational Routines--Part II

Science.gov (United States)

Fiol, C. Marlena; O'Connor, Edward J.

2017-01-01

Purpose: The purpose of Part II of this two-part paper is to uncover important differences in the nature of the three unlearning subprocesses, which call for different leadership interventions to motivate people to move through them. Design/methodology/approach: The paper draws on research in behavioral medicine and psychology to demonstrate that…

Experience with EBR-II [Experimental Breeder Reactor] driver fuel

International Nuclear Information System (INIS)

Seidel, B.R.; Porter, D.L.; Walters, L.C.; Hofman, G.L.

1986-01-01

The exceptional performance of Experimental Breeder Reactor-II (EBR-II) metallic driver fuel has been demonstrated by the irradiation of a large number of elements under steady-state, transient overpower, and loss-of-flow conditions. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. Quantification of reliability has allowed full utilization of element lifetime. Improved design and duct materials currently under test are expected to increase the burnup from 8 to 14 at.%

Systems Analysis of Technologies for Energy Recovery from Waste. Part I. Gasification followed by Catalytic Combustion, PEM Fuel Cells and Solid Oxide Fuel Cells for Stationary Applications in Comparison with Incineration. Part - II. Catalytic combustion - Experimental part

Energy Technology Data Exchange (ETDEWEB)

Assefa, Getachew; Frostell, Bjoern [Royal Inst. of Technology, Stockholm (Sweden). Div. of Industrial Ecology; Jaeraas, Sven; Kusar, Henrik [Royal Inst. of Technology, Stockholm (Sweden). Div. of Chemical Technology

2005-02-01

This project is entitled 'Systems Analysis: Energy Recovery from waste, catalytic combustion in comparison with fuel cells and incineration'. Some of the technologies that are currently developed by researchers at the Royal Institute of Technology include catalytic combustion and fuel cells as downstream units in a gasification system. The aim of this project is to assess the energy turnover as well as the potential environmental impacts of biomass/waste-to-energy technologies. In second part of this project economic analyses of the technologies in general and catalytic combustion and fuel cell technologies in particular will be carried out. Four technology scenarios are studied: (1) Gasification followed by Low temperature fuel cells (Proton Exchange Membrane (PEM) fuel cells) (2) Gasification followed by high temperature fuel cells (Solid Oxide Fuel Cells (SOFC) (3) Gasification followed by catalytic combustion and (4) Incineration with energy recovery. The waste used as feedstock is an industrial waste containing parts of household waste, paper waste, wood residues and poly ethene. In the study compensatory district heating is produced by combustion of biofuel. The power used for running the processes in the scenarios will be supplied by the waste-to-energy technologies themselves while compensatory power is assumed to be produced from natural gas. The emissions from the system studied are classified and characterised using methodology from Life Cycle Assessment in to the following environmental impact categories: Global Warming Potential, Acidification Potential, Eutrophication Potential and finally Formation of Photochemical Oxidants. Looking at the result of the four technology chains in terms of the four impact categories with impact per GWh electricity produced as a unit of comparison and from the perspective of the rank each scenario has in all the four impact categories, SOFC appears to be the winner technology followed by PEM and CC as second

Systems Analysis of Technologies for Energy Recovery from Waste. Part I. Gasification followed by Catalytic Combustion, PEM Fuel Cells and Solid Oxide Fuel Cells for Stationary Applications in Comparison with Incineration. Part - II. Catalytic combustion - Experimental part

International Nuclear Information System (INIS)

Assefa, Getachew; Frostell, Bjoern; Jaeraas, Sven; Kusar, Henrik

2005-02-01

This project is entitled 'Systems Analysis: Energy Recovery from waste, catalytic combustion in comparison with fuel cells and incineration'. Some of the technologies that are currently developed by researchers at the Royal Institute of Technology include catalytic combustion and fuel cells as downstream units in a gasification system. The aim of this project is to assess the energy turnover as well as the potential environmental impacts of biomass/waste-to-energy technologies. In second part of this project economic analyses of the technologies in general and catalytic combustion and fuel cell technologies in particular will be carried out. Four technology scenarios are studied: (1) Gasification followed by Low temperature fuel cells (Proton Exchange Membrane (PEM) fuel cells) (2) Gasification followed by high temperature fuel cells (Solid Oxide Fuel Cells (SOFC) (3) Gasification followed by catalytic combustion and (4) Incineration with energy recovery. The waste used as feedstock is an industrial waste containing parts of household waste, paper waste, wood residues and poly ethene. In the study compensatory district heating is produced by combustion of biofuel. The power used for running the processes in the scenarios will be supplied by the waste-to-energy technologies themselves while compensatory power is assumed to be produced from natural gas. The emissions from the system studied are classified and characterised using methodology from Life Cycle Assessment in to the following environmental impact categories: Global Warming Potential, Acidification Potential, Eutrophication Potential and finally Formation of Photochemical Oxidants. Looking at the result of the four technology chains in terms of the four impact categories with impact per GWh electricity produced as a unit of comparison and from the perspective of the rank each scenario has in all the four impact categories, SOFC appears to be the winner technology followed by PEM and CC as second and third

Fuel penetration of intersubassembly gaps in LMFBRs: a calculational method with the SIMMER-II code

International Nuclear Information System (INIS)

DeVault, G.P.

1983-01-01

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor (LMFBR) undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. A possible avenue for early fuel removal in heterogeneous core LMFBRs is the failure of duct walls in disrupted driver subassemblies followed by fuel penetration into the gaps between blanket subassemblies. The SIMMER-II code was modified to simulate flow between subassembly gaps. Calculations with the modified SIMMER-II code indicate the capabilities of the method and the potential for fuel mass reduction in the active core

Thorium utilization in a small long-life HTR. Part II: Seed-and-blanket fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Ding, Ming, E-mail: dingming@hrbeu.edu.cn [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

2014-02-15

Highlights: • Seed-and-blanket (S and B) fuel blocks are proposed for a small block-type HTR. • S and B fuel blocks consist of a seed region (UO{sub 2}) and a blanket region (ThO{sub 2}). • The neutronic performance of S and B fuel blocks are analyzed using SCALE 6. • Three S and B fuel blocks with a reactivity swing of 0.1 Δk are recommended. • S and B fuel blocks are compared with thorium MOX fuel blocks. - Abstract: In order to utilize thorium in high temperature gas-cooled reactors (HTRs), the concept of seed-and-blanket (S and B) fuel block is introduced into the U-Battery, which is a long-life block-type HTR with a thermal power of 20 MWth. A S and B fuel block consists of a seed region with uranium in the center, and a blanket region with thorium. The neutronic performance, such as the multiplication factor, conversion ratio and reactivity swing, of a typical S and B fuel block was investigated by SCALE 6.0 by parametric analysis of the composition parameters and geometric parameters of the fuel block for the U-Battery application. Since the purpose of U-235 in the S and B fuel block is to ignite the fission reactions in the fuel block, 20% enriched uranium is recommended for the S and B fuel block. When the ratio of the number of carbon to heavy metal atoms changes with the geometric parameters of the fuel block in the range of 200–250, the reactivity swing reaches very small values. Furthermore, for a reactivity swing of 0.1 Δk during 10 effective full power years, three configurations with 36, 54 and 78 UO{sub 2} fuel rods are recommended for the application of the U-Battery. The comparison analysis of the S and B fuel block with the Th/U MOX fuel block shows that the former has a longer lifetime and a lower reactivity swing.

SP-100 Fuel Pin Performance: Results from Irradiation Testing

Science.gov (United States)

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

1994-07-01

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

Fuel Element Technical Manual

Energy Technology Data Exchange (ETDEWEB)

Burley, H.H. [ed.

1956-08-01

It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

Energy Technology Data Exchange (ETDEWEB)

Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable.

X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

International Nuclear Information System (INIS)

Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk

2016-01-01

In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable

Nuclear medicine and thyroid disease - part II

International Nuclear Information System (INIS)

Chatterton, B.E.

2005-01-01

Part 1 of this article discussed the anatomy, physiology and basic pathology of the thyroid gland. Techniques of thyroid scanning and a few clinical examples are shown part II Copyright (2005) The Australian and New Zealand Society Of Nuclear Medicine Inc

Short-stack modeling of degradation in solid oxide fuel cells. Part I. Contact degradation

Energy Technology Data Exchange (ETDEWEB)

Gazzarri, J.I. [Department of Mechanical Engineering, University of British Columbia, 2054-6250 Applied Science Lane, Vancouver, BC V6T 1Z4 (Canada); Kesler, O. [Department of Mechanical and Industrial Engineering, University of Toronto, 5 King' s College Road, Toronto, ON M5S 3G8 (Canada)

2008-01-21

As the first part of a two paper series, we present a two-dimensional impedance model of a working solid oxide fuel cell (SOFC) to study the effect of contact degradation on the impedance spectrum for the purpose of non-invasive diagnosis. The two dimensional modeled geometry includes the ribbed interconnect, and is adequate to represent co- and counter-flow configurations. Simulated degradation modes include: cathode delamination, interconnect oxidation, and interconnect-cathode detachment. The simulations show differences in the way each degradation mode impacts the impedance spectrum shape, suggesting that identification is possible. In Part II, we present a sensitivity analysis of the results to input parameter variability that reveals strengths and limitations of the method, as well as describing possible interactions between input parameters and concurrent degradation modes. (author)

Short-stack modeling of degradation in solid oxide fuel cells. Part I. Contact degradation

Science.gov (United States)

Gazzarri, J. I.; Kesler, O.

As the first part of a two paper series, we present a two-dimensional impedance model of a working solid oxide fuel cell (SOFC) to study the effect of contact degradation on the impedance spectrum for the purpose of non-invasive diagnosis. The two dimensional modeled geometry includes the ribbed interconnect, and is adequate to represent co- and counter-flow configurations. Simulated degradation modes include: cathode delamination, interconnect oxidation, and interconnect-cathode detachment. The simulations show differences in the way each degradation mode impacts the impedance spectrum shape, suggesting that identification is possible. In Part II, we present a sensitivity analysis of the results to input parameter variability that reveals strengths and limitations of the method, as well as describing possible interactions between input parameters and concurrent degradation modes.

Time Evolution of Selected Actinides in TRIGA MARK-II Fuel

International Nuclear Information System (INIS)

Usang, M.D.; Naim Shauqi Hamzah; Mohamad Hairie Rabir

2011-01-01

Study is made on the evolution of several actinides capable of undergoing fission or breeding available on the Malaysian Nuclear Agency (MNA) TRIGA MARK-II fuel. Population distribution of burned fuel in the MNA reactor is determined with a model developed using WIMS. This model simulates fuel conditions in the hottest position in the reactor, thus the location where most of the burn up occurs. Theoretical basis of these nuclide time evolution are explored and compared with the population obtained from our models. Good agreements are found for the theoretical time evolution and the population of Uranium-235, Uranium-236, Uranium-238 and Plutonium-239. (author)

Fuels and fire in land-management planning. Part 1. Forest-fuel classification.

Science.gov (United States)

Wayne G. Maxwell; Franklin R. Ward

1981-01-01

This report describes a way to collect and classify the total fuel complex within a planning area. The information can be used as input for appraising and rating probable fire behavior and calculating expected costs and losses from various land uses and management alternatives, reported separately as Part 2 and Part 3 of this series. This total package can be used...

Power and power-to-flow reactivity transfer functions in EBR-II [Experimental Breeder Reactor II] fuel

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1989-01-01

Reactivity transfer functions are important in determining the reactivity history during a power transient. Overall nodal transfer functions have been calculated for different subassembly types in the Experimental Breeder Reactor II (EBR-II). Steady-state calculations for temperature changes and, hence, reactivities for power changes have been separated into power and power-to-flow-dependent terms. Axial nodal transfer functions separated into power and power-to-flow-dependent components are reported in this paper for a typical EBR-II fuel pin. This provides an improved understanding of the time dependence of these components in transient situations

Fuel Management Strategies for a Possible Future LEU Core of a TRIGA Mark II Vienna

Energy Technology Data Exchange (ETDEWEB)

Khan, R.; Villa, M.; Steinhauser, G.; Boeck, H. [Vienna University of Technology-Atominstitut (Austria)

2011-07-01

The Vienna University of Technology/Atominstitut (VUT/ATI) operates a TRIGA Mark II research reactor. It is operated with a completely mixed core of three different types of fuel. Due to the US fuel return program, the ATI have to return its High Enriched Uranium (HEU) fuel latest by 2019. As an alternate, the Low Enrich Uranium (LEU) fuel is under consideration. The detailed results of the core conversion study are presented at the RRFM 2011 conference. This paper describes the burn up calculations of the new fuel to predict the future burn up behavior and core life time. It also develops an effective and optimized fuel management strategy for a possible future operation of the TRIGA Mark II with a LEU core. This work is performed by the combination of MCNP5 and diffusion based neutronics code TRIGLAV. (author)

Recent Economic Perspectives on Political Economy, Part II*

Science.gov (United States)

Dewan, Torun; Shepsle, Kenneth A.

2013-01-01

In recent years some of the best theoretical work on the political economy of political institutions and processes has begun surfacing outside the political science mainstream in high quality economics journals. This two-part paper surveys these contributions from a recent five-year period. In Part I, the focus is on elections, voting and information aggregation, followed by treatments of parties, candidates, and coalitions. In Part II, papers on economic performance and redistribution, constitutional design, and incentives, institutions, and the quality of political elites are discussed. Part II concludes with a discussion of the methodological bases common to economics and political science, the way economists have used political science research, and some new themes and arbitrage opportunities. PMID:23606754

Potential use of California lignite and other alternate fuel for enhanced oil recovery. Phase I and II. Final report. [As alternative fuels for steam generation in thermal EOR

Energy Technology Data Exchange (ETDEWEB)

Shelton, R.; Shimizu, A.; Briggs, A.

1980-02-01

The Nation's continued reliance on liquid fossil fuels and decreasing reserves of light oils gives increased impetus to improving the recovery of heavy oil. Thermal enhanced oil recovery EOR techniques, such as steam injection, have generally been the most effective for increasing heavy oil production. However, conventional steam generation consumes a large fraction of the produced oil. The substitution of alternate (solid) fuels would release much of this consumed oil to market. This two-part report focuses on two solid fuels available in California, the site of most thermal EOR - petroleum coke and lignite. Phase I, entitled Economic Analysis, shows detailed cost comparisons between the two candidate fuels and also with Western coal. The analysis includes fuels characterizations, process designs for several combustion systems, and a thorough evaluation of the technical and economic uncertainties. In Phase II, many technical parameters of petroleum coke combustion were measured in a pilot-plant fluidized bed. The results of the study showed that petroleum coke combustion for EOR is feasible and cost effective in a fluidized bed combustor.

Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

1988-01-01

The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

Development of metal fuel and study of construction materials (I-IV), Part II

International Nuclear Information System (INIS)

Mihajlovic, A.

1965-11-01

The studies were devoted to problems related to application of metal uranium as fuel in heavy water reactors. Influence of thermal treatment on material texture and recrystallization of cast uranium was investigated. Structural changes of uranium alloys with molybdenum and niobium were tested during different heat treatments. A review of the possibilities for using metal uranium fuel in heavy water reactors is included

Improvements in fabrication of metallic fuels

International Nuclear Information System (INIS)

Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

1989-12-01

Argonne National Laboratory is currently developing a new liquid- metal cooled breeder reactor known as the Integral Fast Reactor (IFR). IFR fuels represent the state-of-the-art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, to be discussed below, will support the fully remote fuel cycle facility that as an integral part of the IFR concept will be demonstrated at the EBR-II site. 3 refs

Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

1993-01-01

Three furnace heating tests were conducted with irradiated, HT9-clad and U-19wt%Pu-10wt%Zr-alloy, EBR-II Mk-V-type fuel elements to evaluate the behavior that could be expected during a loss-of-flow event in the reactor. In general, very significant safety margins for cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results are presented, as are discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction that were found in these tests. (orig.)

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

Science.gov (United States)

2010-01-01

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

COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

International Nuclear Information System (INIS)

Rhodes, C.A.

1984-12-01

This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

Review of the micro-tubular solid oxide fuel cell. Part I. Stack design issues and research activities

Energy Technology Data Exchange (ETDEWEB)

Lawlor, V. [Department of Eco-Energy Engineering, Upper Austrian University of Applied Sciences, A-4600 Wels (Austria); Department of Manufacturing and Mechanical Engineering, Dublin City University, Dublin 9 (Ireland); Griesser, S. [Department of Eco-Energy Engineering, Upper Austrian University of Applied Sciences, A-4600 Wels (Austria); Buchinger, G. [eZelleron GmbH, Collenbusch str. 22, 01324 Dresden (Germany); Olabi, A.G. [Department of Manufacturing and Mechanical Engineering, Dublin City University, Dublin 9 (Ireland); Cordiner, S. [Dipartimento di Ingegneria Meccanica - Universita di Roma Tor Vergata (Italy); Meissner, D. [Department of Eco-Energy Engineering, Upper Austrian University of Applied Sciences, A-4600 Wels (Austria); Department of Material Science, Tallinn University of Technology, Ehitajate 19086 (Estonia)

2009-09-05

Fuel cells are devices that convert chemical energy in hydrogen enriched fuels into electricity electrochemically. Micro-tubular solid oxide fuel cells (MT-SOFCs), the type pioneered by K. Kendall in the early 1990s, are a variety of SOFCs that are on the scale of millimetres compared to their much larger SOFC relatives that are typically on the scale of tens of centimetres. The main advantage of the MT-SOFC, over its larger predecessor, is that it is smaller in size and is more suitable for rapid start up. This may allow the SOFC to be used in devices such as auxiliary power units, automotive power supplies, mobile electricity generators and battery re-chargers. The following paper is Part I of a two part series. Part I will introduce the reader to the MT-SOFC stack and its applications, indicating who is researching what in this field and also specifically investigate the design issues related to multi-cell reactor systems called stacks. Part II will review in detail the combinations of materials and methods used to produce the electrodes and electrolytes of MT-SOFC's. Also the role of modelling and validation techniques used in the design and improvement of the electrodes and electrolytes will be investigated. A broad range of scientific and engineering disciplines are involved in a stack design. Scientific and engineering content has been discussed in the areas of thermal-self-sustainability and efficiency, sealing technologies, manifold design, electrical connections and cell performance optimisation. (author)

Fossil fuels: Kyoto initiatives and opportunities. Part 1

International Nuclear Information System (INIS)

Pinelli, G.; Zerlia, T.

2008-01-01

GHG emission in the upstream step of fossil fuel chains could give an environmental as well as economic opportunity for traditional sectors. This study deepens the matter showing an increasing number of initiative over the last few years taken both the involved sectors and by various stake holders (public and private subjects) within the Kyoto flexible mechanism (CDM and JI) or linked to voluntary national or at a global level actions. The above undertakings give evidence for an increased interest and an actual activity dealing with GHG reduction whose results play an evident and positive role for the environment too. Part 1. of this study deals with fossil fuel actions within the Kyoto protocol mechanism. Part 2. will show international and national voluntary initiative [it

Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

International Nuclear Information System (INIS)

Chang, L.K.; Koenig, J.F.; Porter, D.L.

1987-01-01

In the Shutdown Heat Removal Testing (SHRT) program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elemnts, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date. (orig.)

Short stack modeling of degradation in solid oxide fuel cells. Part II. Sensitivity and interaction analysis

Science.gov (United States)

Gazzarri, J. I.; Kesler, O.

In the first part of this two-paper series, we presented a numerical model of the impedance behaviour of a solid oxide fuel cell (SOFC) aimed at simulating the change in the impedance spectrum induced by contact degradation at the interconnect-electrode, and at the electrode-electrolyte interfaces. The purpose of that investigation was to develop a non-invasive diagnostic technique to identify degradation modes in situ. In the present paper, we appraise the predictive capabilities of the proposed method in terms of its robustness to uncertainties in the input parameters, many of which are very difficult to measure independently. We applied this technique to the degradation modes simulated in Part I, in addition to anode sulfur poisoning. Electrode delamination showed the highest robustness to input parameter variations, followed by interconnect oxidation and interconnect detachment. The most sensitive degradation mode was sulfur poisoning, due to strong parameter interactions. In addition, we simulate several simultaneous two-degradation-mode scenarios, assessing the method's capabilities and limitations for the prediction of electrochemical behaviour of SOFC's undergoing multiple simultaneous degradation modes.

Short stack modeling of degradation in solid oxide fuel cells. Part II. Sensitivity and interaction analysis

Energy Technology Data Exchange (ETDEWEB)

Gazzarri, J.I. [Department of Mechanical Engineering, University of British Columbia, 2054-6250 Applied Science Lane, Vancouver, BC V6T 1Z4 (Canada); Kesler, O. [Department of Mechanical and Industrial Engineering, University of Toronto, 5 King' s College Road, Toronto, ON M5S 3G8 (Canada)

2008-01-21

In the first part of this two-paper series, we presented a numerical model of the impedance behaviour of a solid oxide fuel cell (SOFC) aimed at simulating the change in the impedance spectrum induced by contact degradation at the interconnect-electrode, and at the electrode-electrolyte interfaces. The purpose of that investigation was to develop a non-invasive diagnostic technique to identify degradation modes in situ. In the present paper, we appraise the predictive capabilities of the proposed method in terms of its robustness to uncertainties in the input parameters, many of which are very difficult to measure independently. We applied this technique to the degradation modes simulated in Part I, in addition to anode sulfur poisoning. Electrode delamination showed the highest robustness to input parameter variations, followed by interconnect oxidation and interconnect detachment. The most sensitive degradation mode was sulfur poisoning, due to strong parameter interactions. In addition, we simulate several simultaneous two-degradation-mode scenarios, assessing the method's capabilities and limitations for the prediction of electrochemical behaviour of SOFC's undergoing multiple simultaneous degradation modes. (author)

On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

International Nuclear Information System (INIS)

Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

1976-05-01

The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Mariani, R.D.

1999-01-01

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr alloy fuel elements irradiated in the experimental breeder reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining. (orig.)

Fuel and fuel pin behaviour in a high burnup fast breeder fuel subassembly: Results of destructive post-irradiation examinations of the KNK II/1 fuel subassembly NY-205

International Nuclear Information System (INIS)

Patzer, G.

1991-05-01

The report gives a summarizing overview of the design characteristics, of the irradiation history and of the results of the destructive post-irradiation examinations of the fuel pins of the high-burnup fuel subassembly NY-205 of the KNK II first core. This element was operated for about 10 years and reached a maximum local burnup of 175 MWd/kg(HM) and a maximum neutron dose of 67 dpa-NRT. The main design data of this subassembly agree with those of the SNR 300 Mark-Ia, and it reached more than twice of the burnup and a similar neutron dose as foreseen for the SNR 300 fuel subassemblies [de

Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.; Billone, M.C.; Kramer, J.M.

1992-01-01

The next step in the development of metal fuels for the integral fast reactor (IFR) is the conversion of the Experimental Breeder Reactor II (EBR-II) core to one containing the ternary U-20 Pu-10 Zr alloy clad with HT-9 cladding, i.e., the Mk-V core. This paper presents results of three hot-cell furnace simulation tests on irradiated Mk-V-type fuel elements (U-19 Pu-10 Zr/HT-9), which were performed to support the safety case for the Mk-V core. These tests were designed to envelop an umbrella (bounding) unlikely loss-of-flow (LOF) event in EBR-II during which the calculated peak cladding temperature would reach 776 degree C for < 2 min. The principal objectives of these tests were (a) demonstration of the safety margin of the fuel element, (b) investigation of cladding breaching behavior, and (c) provision of data for validation of the FPIN2 and LIFE-METAL codes

Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

International Nuclear Information System (INIS)

Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

1982-01-01

The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling

10 CFR Appendix to Part 474 - Sample Petroleum-Equivalent Fuel Economy Calculations

Science.gov (United States)

2010-01-01

... 10 Energy 3 2010-01-01 2010-01-01 false Sample Petroleum-Equivalent Fuel Economy Calculations..., DEVELOPMENT, AND DEMONSTRATION PROGRAM; PETROLEUM-EQUIVALENT FUEL ECONOMY CALCULATION Pt. 474, App. Appendix to Part 474—Sample Petroleum-Equivalent Fuel Economy Calculations Example 1: An electric vehicle is...

40 CFR Appendix Viii to Part 600 - Fuel Economy Label Formats

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Fuel Economy Label Formats VIII... POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Pt. 600, App. VIII Appendix VIII to Part 600—Fuel Economy Label Formats EC01MY92.117 EC01MY92.118 EC01MY92.119 EC01MY92.120...

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-01-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

Energy Technology Data Exchange (ETDEWEB)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-01-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

Alternate-Fueled Combustor-Sector Performance. Parts A and B; (A) Combustor Performance; (B) Combustor Emissions

Science.gov (United States)

Shouse, D. T.; Hendricks, R. C.; Lynch, A.; Frayne, C. W.; Stutrud, J. S.; Corporan, E.; Hankins, T.

2012-01-01

Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as "drop-in" fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of processing and assumed to be feedstock agnostic. Adherence to alternate fuels and fuel blends requires "smart fueling systems" or advanced fuel-flexible systems, including combustors and engines, without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data. The data are for nominal inlet conditions at 225 psia and 800 F (1.551 MPa and 700 K), for synthetic-paraffinic-kerosene- (SPK-) type (Fisher-Tropsch (FT)) fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling. Assessments are made of the change in combustor efficiency, wall temperatures, emissions, and luminosity with SPK of 0%, 50%, and 100% fueling composition at 3% combustor pressure drop. The performance results (Part A) indicate no quantifiable differences in combustor efficiency, a general trend to lower liner and higher core flow temperatures with increased FT fuel blends. In general, emissions data (Part B) show little differences, but with percent increase in FT-SPK-type fueling, particulate emissions and wall temperatures are less than with baseline JP-8. High-speed photography illustrates both luminosity and combustor dynamic flame characteristics.

IMPROVEMENT OF PERFORMANCE OF DUAL FUEL ENGINE OPERATED AT PART LOAD

Directory of Open Access Journals (Sweden)

N. Kapilan

2010-12-01

Full Text Available Rising petroleum prices, an increasing threat to the environment from exhaust emissions, global warming and the threat of supply instabilities has led to the choice of inedible Mahua oil (MO as one of the main alternative fuels to diesel oil in India. In the present work, MO was converted into biodiesel by transesterification using methanol and sodium hydroxide. The cost of Mahua oil biodiesel (MOB is higher than diesel. Hence liquefied petroleum gas (LPG, which is one of the cheapest gaseous fuels available in India, was fumigated along with the air to reduce the operating cost and to reduce emissions. The dual fuel engine resulted in lower efficiency and higher emissions at part load. Hence in the present work, the injection time was varied and the performance of the dual fuel engine was studied. From the engine tests, it is observed that an advanced injection time results in higher efficiency and lower emissions. Hence, advancing the injection timing is one of the ways of increasing the efficiency of LPG+MOB dual fuel engine operated at part load.

40 CFR Appendix III to Part 600 - Sample Fuel Economy Label Calculation

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Sample Fuel Economy Label Calculation...) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Pt. 600, App. III Appendix III to Part 600—Sample Fuel Economy Label Calculation Suppose that a manufacturer called Mizer...

Irradiation of a 19 pin subassembly with mixed carbide fuel in KNK II

Science.gov (United States)

Geithoff, D.; Mühling, G.; Richter, K.

1992-06-01

The presentation deals with the fabrication, irradiation and nondestructive postirradiation examinations of LMR fuel pins with mixed (U, Pu)-carbide fuels. The mixed carbide fuel was fabricated by the European Institute of Transuranium Elements using various fabrication procedures. Fuel composition varied therefore in a wide range of tolerances with respect to oxygen and phase content and microstructure. The 19 carbide pins were irradiated in the fast neutron flux of the KNK II reactor to a burn-up of about 7 at% without any failure in the centre of a KNK "carrier element" at a maximum linear rating of 800 W/cm. After dismantling in the Hot Cells of KfK nondestructive examinations were carried out comprising dimensional controls, radiography, γ-scanning and eddy-current testing. The results indicate differences in fuel behaviour with respect to composition of the fuel.

PIO I-II tendencies. Part 2. Improving the pilot modeling

Directory of Open Access Journals (Sweden)

Ioan URSU

2011-03-01

Full Text Available The study is conceived in two parts and aims to get some contributions to the problem ofPIO aircraft susceptibility analysis. Part I, previously published in this journal, highlighted the mainsteps of deriving a complex model of human pilot. The current Part II of the paper considers a properprocedure of the human pilot mathematical model synthesis in order to analyze PIO II typesusceptibility of a VTOL-type aircraft, related to the presence of position and rate-limited actuator.The mathematical tools are those of semi global stability theory developed in recent works.

Fuel burnup analysis of the TRIGA Mark II reactor at the University of Pavia

International Nuclear Information System (INIS)

Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

2016-01-01

Highlights: • A fuel evolution model for a TRIGA Mark II reactor has been developed. • Reproduction of nearly 50 years of reactor operation. • The model was used to predict the best reactor reconfiguration. • Reactor life was extended without adding fresh fuel elements. - Abstract: A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyze neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate low power experimental reactors from those used for power production, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the core and to obtain a substantial increase in the Core Excess reactivity value. The evaluation of fuel burnup and the reconfiguration results are presented in this paper.

System modelling to support accelerated fuel transfer rate at EBR-II

International Nuclear Information System (INIS)

Imel, G.R.; Houshyar, A.; Planchon, H.P.; Cutforth, D.C.

1995-01-01

The Experimental Breeder Reactor-II (EBR-II) ia a 62.5 MW(th) liquid metal reactor operated by Argonne National Laboratory for The United States Department of Energy. The reactor is located near Idaho Falls, Idaho at the Argonne-West site (ANL-W). Full power operation was achieved in 1964,- the reactor operated continuously since that time until October 1994 in a variety of configurations depending on the programmatic mission. A three year program was initiated in October, 1993 to replace the 330 depleted uranium blanket subassemblies (S/As) with stainless steel reflectors. It was intended to operate the reactor during the three year blanket unloading program, followed by about a half year of driver fuel unloading. However, in the summer of 1994, Congress dictacted that EBR-II be shut down October 1, and complete defueling without operation. To assist in the planning for resources needed for this defueling campaign, a mathematical model of the fuel handling sequence was developed utilizing the appropriate reliability factors and inherent mm constraints of each stage of the process. The model allows predictions of transfer rates under different scenarios. Additionally, it has facilitated planning of maintenance activities, as well as optimization of resources regarding manpower and modification effort. The model and its application is described in this paper

Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1962-07-01

Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U 235 chain, analytical expressions for the concentrations of U 235 , U 236 and Np 237 as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer

Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

1992-11-01

This report discusses three furnace heating tests which were conducted with irradiated, HT9-clad and U-19wt.%Pu-l0wt.%Zr-alloy fuel, Mk-V-type fuel elements in the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. In general, very significant safety margins for fuel-element cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results will be given, as well as discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction found in high-temperature testing of irradiated metallic fuel elements

Validating the standard for the National Board Dental Examination Part II.

Science.gov (United States)

Tsai, Tsung-Hsun; Neumann, Laura M; Littlefield, John H

2012-05-01

As part of the overall exam validation process, the Joint Commission on National Dental Examinations periodically reviews and validates the pass/fail standard for the National Board Dental Examination (NBDE), Parts I and II. The most recent standard-setting activities for NBDE Part II used the Objective Standard Setting method. This report describes the process used to set the pass/fail standard for the 2009 exam. The failure rate on the NBDE Part II increased from 5.3 percent in 2008 to 13.7 percent in 2009 and then decreased to 10 percent in 2010. This article describes the Objective Standard Setting method and presents the estimated probabilities of classification errors based on the beta binomial mathematical model. The results show that the probability of correct classifications of candidate performance is very high (0.97) and that probabilities of false negative and false positive errors are very small (.03 and <0.001, respectively). The low probability of classification errors supports the conclusion that the pass/fail score on the NBDE Part II is a valid guide for making decisions about candidates for dental licensure.

EBR-II argon cooling system restricted fuel handling I and C upgrade

International Nuclear Information System (INIS)

Start, S.E.; Carlson, R.B.; Gehrman, R.L.

1995-01-01

The instrumentation and control of the Argon Cooling System (ACS) restricted fuel handling control system at Experimental Breeder Reactor II (EBR-II) is being upgraded from a system comprised of many discrete components and controllers to a computerized system with a graphical user interface (GUI). This paper describes the aspects of the upgrade including reasons for the upgrade, the old control system, upgrade goals, design decisions, philosophies and rationale, and the new control system hardware and software

A transient overpower experiment in EBR-II

International Nuclear Information System (INIS)

Herzog, J.P.; Tsai, H.; Dean, E.M.; Aoyama, T.; Yamamoto, K.

1994-01-01

The TOPI-IE test was a transient overpower test on irradiate mixed-oxide fuel pins in the Experimental Breeder Reactor-II (EBR-II). The test, the fifth in a series, was part of a cooperative program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan to conduct operational transient testing on mixed-oxide fuel pins in the metal-fueled EBR-II. The principle objective of the TOPI-1E test was to assess breaching margins for irradiated mixed-oxide fuel pins over the Plant Protection System (PPS) thresholds during a slow, extended overpower transient. This paper describes the effect of the TOPI-1E experiment on reactor components and the impact of the experiment on the long-term operability of the reactor. The paper discusses the role that SASSYS played in the pre-test safety analysis of the experiment. The ability of SASSYS to model transient overpower events is detailed by comparisons of data from the experiment with computed reactor variables from a SASSYS post-test simulation of the experiment

AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XXIII, I--MAINTAINING THE FUEL SYSTEM, PART II--CATERPILLAR DIESEL ENGINE, II--UNDERSTANDING STEERING SYSTEMS.

Science.gov (United States)

Minnesota State Dept. of Education, St. Paul. Div. of Vocational and Technical Education.

THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL INJECTION SYSTEM AND THE STEERING SYSTEM OF DIESEL POWERED VEHICLES. TOPICS ARE FUEL INJECTION SECTION, AND DESCRIPTION OF THE STEERING SYSTEM. THE MODULE CONSISTS OF A SELF-INSTRUCTIONAL BRANCH PROGRAMED TRAINING…

Evolution of thermal-hydraulics testing in EBR-II

International Nuclear Information System (INIS)

Golden, G.H.; Planchon, H.P.; Sackett, J.I.; Singer, R.M.

1987-01-01

A thermal-hydraulics testing and modeling program has been underway at the Experimental Breeder Reactor-II (EBR-II) for 12 years. This work culminated in two tests of historical importance to commercial nuclear power, a loss of flow without scram and a loss of heat sink wihout scram, both from 100% initial power. These tests showed that natural processes will shut EBR-II down and maintain cooling without automatic control rod action or operator intervention. Supporting analyses indicate that these results are characteristic of a range of sizes of liquid metal cooled reactors (LMRs), if these reactors use metal driver fuel. This type of fuel is being developed as part of the Integral Fast Reactor Program at Argonne National Laboratory. Work is now underway at EBR-II to exploit the inherent safety of metal-fueled LMRs with regard to development of improved plant control strategies. (orig.)

Behaviour of gas cooled reactor fuel under accident conditions

International Nuclear Information System (INIS)

1991-11-01

The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

Globalization in the pharmaceutical industry, Part II.

Science.gov (United States)

Casadio Tarabusi, C; Vickery, G

1998-01-01

This is the second of a two-part report on the pharmaceutical industry. Part II begins with a discussion of foreign direct investment and inter-firm networks, which covers international mergers, acquisitions, and minority participation; market shares of foreign-controlled firms; international collaboration agreements (with a special note on agreements in biotechnology); and licensing agreements. The final section of the report covers governmental policies on health and safety regulation, price regulation, industry and technology, trade, foreign investment, protection of intellectual property, and competition.

Mechanical behaviour of a fuel cell stack under vibrating conditions linked to aircraft applications part II: Three-dimensional modelling

Energy Technology Data Exchange (ETDEWEB)

Rouss, Vicky; Charon, Willy [M3M, University of Technology Belfort - Montbeliard (France); FCLAB, Rue Thierry Mieg, F 90010 Belfort, Cedex (France); Candusso, Denis [INRETS, The French National Institute for Transport and Safety Research (France); FCLAB, Rue Thierry Mieg, F 90010 Belfort, Cedex (France)

2008-11-15

The implementation of fuel cells (FC) in transportation systems such as airplanes requires better understanding of their mechanical behaviour in vibrating environment. To this end, a FC stack was tested on a vibrating platform for all three orthogonal axes. The experimental procedure is described in the first part of the paper. This second part of the paper demonstrates how the experimental data collected can be used to create a three-dimensional, multi-input and multi-output model based on the Artificial Neural Network (ANN) approach. Indeed FCs are nonlinear mechanical systems, difficult to be physically modelled. The ANN methodology which depends strictly on raw data is a particularly interesting alternative solution to model FCs, for example, for monitoring purpose. The ANN model is described along with the training, pruning and validation stages. The results are exposed and commented. (author)

29 CFR Appendix II to Part 1918 - Tables for Selected Miscellaneous Auxiliary Gear (Mandatory)

Science.gov (United States)

2010-07-01

... 29 Labor 7 2010-07-01 2010-07-01 false Tables for Selected Miscellaneous Auxiliary Gear (Mandatory) II Appendix II to Part 1918 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND.... 1918, App. II Appendix II to Part 1918—Tables for Selected Miscellaneous Auxiliary Gear (Mandatory...

Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

International Nuclear Information System (INIS)

DeVault, G.P.; Bell, C.R.

1985-01-01

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

International Nuclear Information System (INIS)

Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

2008-08-01

The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

HANARO fuel irradiation test(II)

Energy Technology Data Exchange (ETDEWEB)

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

2001-04-01

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

Mission Plan for the Civilian Radioactive Waste Management Program. Volume I. Part I. Overview and current program plans; Part II. Information required by the Nuclear Waste Policy Act of 1982

International Nuclear Information System (INIS)

1985-06-01

The Misson Plan is divided into two parts. Part I describes the overall goals, objectives, and strategy for the disposal of spent nuclear fuel and high-level waste. It explains that, to meet the directives of the Nuclear Waste Policy Act, the DOE intends to site, design, construct, and start operating a mined geologic repository by January 31, 1998. The Act specifies that the costs of these activities will be borne by the owners and generators of the waste received at the repository. Part I further describes the other components of the waste-management program - monitored retrievable storage, Federal interim storage, and transportation - as well as systems integration activities. Also discussed are institutional plans and activities as well as the program-management system being implemented by the Office of Civilian Radioactive Waste Management. Part II of the Mission Plan presents the detailed information required by Section 301(a) of the Act - key issues and information needs; plans for obtaining the necessary information; potential financial, institutional, and legal issues; plans for the test and evaluation facility; the principal results obtained to date from site investigations; information on the site-characterization programs; information on the waste package; schedules; costs; and socioeconomic impacts. In accordance with Section 301(a) of the Act, Part II is concerned primarily with the repository program

46 CFR Table II to Part 150 - Grouping of Cargoes

Science.gov (United States)

2010-10-01

... solution Potassium oleate Potassium salt of polyolefin acid Propyl acetate Propylene carbonate Propylene... lignosulfonate solution Sodium polyacrylate solution 2 Sodium salt of Ferric hydroxyethylethylenediamine... 46 Shipping 5 2010-10-01 2010-10-01 false Grouping of Cargoes II Table II to Part 150 Shipping...

Vibrational effects of fuel elements detected during KNK II power operation

International Nuclear Information System (INIS)

Mitzel, F.; Vaeth, W.; Ansari, S.

1982-08-01

The reactivity signal of the KNK II reactor shows almost harmonic reactivity oscillations of Δρ≤0.5 cent. Sensitive correlation measurements, made during the regular plant operation with the normal out-of-core plant instrumentation, revealed that they are associated with individual fuel elements. Auxiliary measurements under various operational conditions and theoretical considerations showed that the oscillations are caused by flow-induced mechanical vibrations. Similar characteristics with respect to the frequencies of these oscillations have obviously not yet been observed for fuel element vibrations in other reactors and tests in out-of-core loops. Therefore efforts were made to classify the phenomenon and to identify the excitation mechanism by using only the normal plant instrumentation. It seems to be most likely a flow-induced vibration of whole fuel elements by vortex shedding or jet switching. This model can explain all observations without exception [de

Fuel temperature influence on the performance of a last generation common-rail diesel ballistic injector. Part II: 1D model development, validation and analysis

International Nuclear Information System (INIS)

Payri, R.; Salvador, F.J.; Carreres, M.; De la Morena, J.

2016-01-01

Highlights: • A 1D model of a solenoid common-rail ballistic injector is implemented in AMESim. • A detailed dimensional and a hydraulic characterization lead to a fair validation. • Fuel temperature influence on injector dynamics is assessed through 1D simulations. • Temperature impacts through changes in inlet orifice regime and viscous friction. • Cold fuel temperature leads to a slower injection opening due to high viscosity. - Abstract: A one-dimensional model of a solenoid-driven common-rail diesel injector has been developed in order to study the influence of fuel temperature on the injection process. The model has been implemented after a thorough characterization of the injector, both from the dimensional and the hydraulic point of view. In this sense, experimental tools for the determination of the geometry of the injector lines and orifices have been described in the paper, together with the hydraulic setup introduced to characterize the flow behaviour through the calibrated orifices. An extensive validation of the model has been performed by comparing the modelled mass flow rate against the experimental results introduced in the first part of the paper, which were performed for different engine-like operating conditions involving a wide range of fuel temperatures, injection pressures and energizing times. In that first part of the study, an important influence of the fuel temperature was reported, especially in terms of the dynamic behaviour of the injector, due to its ballistic nature. The results from the model have allowed to explain and further extend the findings of the experimental study by analyzing key features of the injector dynamics, such as the pressure drop established in the control volume due to the control orifices performance or the forces due to viscous friction, also assessing their influence on the needle lift laws.

Typewriting Syllabus: Part II: Modules. 1976 Revision.

Science.gov (United States)

New York State Education Dept., Albany. Bureau of Occupational and Career Curriculum Development.

The document is the second of a two-part set on typewriting and focuses on the nine modules of instruction. The nine modules are: (1) keyboard mastery and skill development, (2) basic typewriting competencies, (2a) personal use typewriting, (3) introduction to office typewriting I, (4) introduction to office typewriting II, (5) intermediate office…

Fuel depletion analyses for the HEU core of GHARR-1: Part II: Fission product inventory

International Nuclear Information System (INIS)

Anim-Sampong, S.; Akaho, E.H.K.; Boadu, H.O.; Intsiful, J.D.K.; Osae, S.

1999-01-01

The fission product isotopic inventories have been estimated for a 90.2% highly enriched uranium (HEU) fuel lattice cell of the Ghana Research Reactor-1 (GHARR-1) using the WIMSD/4 transport lattice code. The results indicate a gradual decrease in the Xe 135 inventory, and saturation trend for Sm 149 , Cs 134 and Cs 135 inventories as the fuel is depleted to 10,000 MWd/tU. (author)

Metallographic examinations of the wear-marks on fuel pins of the KNK II/2 fuel assembly NY-308

International Nuclear Information System (INIS)

Patzer, G.

1987-12-01

On the fuel pins and pin spacers of the fuel assembly NY-308 of the second core of KNK II pronounced wear marks had been found in the area of the contact points. In order to determine the exact form of the marks, metallographic investigations were performed on two test pieces of fuel pins in the Hot Cells of the KfK Karlsruhe. It was found that the wear marks did show the already observed stratified structure. Next to the unchanged cladding area there is a peripheral zone with modified grain structure, followed by a layer of moved material and finally there is a flake-like zone of accumulated cladding material at the lower end of the wear marks. Longitudinal cuts do not show grain deformations, which could indicate axial friction forces between pin and spacer. The wear marks are rapidly dropping to their maximum depth at the ends and the depth shows a relatively uniform pattern between both. The findings are confirming the picture, that a stirring movement of the fuel pins took place, which caused adhesive wear [de

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

International Nuclear Information System (INIS)

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

2000-01-01

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

46 CFR Appendix II to Part 150 - Explanation of Figure 1

Science.gov (United States)

2010-10-01

... COMPATIBILITY OF CARGOES Pt. 150, App. II Appendix II to Part 150—Explanation of Figure 1 Definition of a..., aromatic hydrocarbons or paraffins. Others will form hazardous combinations with many groups: For example...

Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

2000-04-01

As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed.

Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

International Nuclear Information System (INIS)

Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

2000-04-01

As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed

Evaluation of the ceramographies of the KNK II/1 test zone fuel assembly NY-202-IA

International Nuclear Information System (INIS)

Geier, F.

1983-12-01

From the 211 fuel pins of the KNK II/1 fuel assembly NY-202-IA six intact fuel pins were selected in addition to the defective pin for destructive post-irradiation examinations in the Hot Cells of the KfK Karlsruhe. The assembly had been unloaded due to a pin failure after 192 equivalent full-power days and a maximum burnup of 5.4 %. The main aspect of these investigations was to record the fuel and fuel pin behavior and thus to allow a comparison of the status before and after irradiation. The results can also be used for comparative calculations and adaptations of existing calculational models. This report documents in detailed form the results of the fuel and fuel pin examinations [de

40 CFR Appendix II to Part 1042 - Steady-State Duty Cycles

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Steady-State Duty Cycles II Appendix..., App. II Appendix II to Part 1042—Steady-State Duty Cycles (a) The following duty cycles apply as specified in § 1042.505(b)(1): (1) The following duty cycle applies for discrete-mode testing: E3 mode No...

Experience gathered from the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type

International Nuclear Information System (INIS)

Pastorini, A.; Belinco, C.G.; El Bis, E.D.; Sacchi, M.A.; Mayans, C.O.; Martin Ghiselli, A.; Marcora, G.R.

1990-01-01

This work describes the needs to materialize the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type, under special conditions, from the Fabrication Pilot Plant sited at the Constituyentes Atomic Center and the Ezeiza Atomic Center, for its subsequent analysis at the High Pressure Experimental Loop. The special conditions under which the transport has been made responded to the fact that the prototype presents a fragile adjustment between rods and separators, necessary to be preserved. (Author) [es

Kick, Glide, Pole! Cross-Country Skiing Fun (Part II)

Science.gov (United States)

Duoos, Bridget A.

2012-01-01

Part I of Kick, Glide, Pole! Cross-Country Skiing Fun, which was published in last issue, discussed how to select cross-country ski equipment, dress for the activity and the biomechanics of the diagonal stride. Part II focuses on teaching the diagonal stride technique and begins with a progression of indoor activities. Incorporating this fun,…

PREREM: an interactive data preprocessing code for INREM II. Part I: user's manual. Part II: code structure

Energy Technology Data Exchange (ETDEWEB)

Ryan, M.T.; Fields, D.E.

1981-05-01

PREREM is an interactive computer code developed as a data preprocessor for the INREM-II (Killough, Dunning, and Pleasant, 1978a) internal dose program. PREREM is intended to provide easy access to current and self-consistent nuclear decay and radionuclide-specific metabolic data sets. Provision is made for revision of metabolic data, and the code is intended for both production and research applications. Documentation for the code is in two parts. Part I is a user's manual which emphasizes interpretation of program prompts and choice of user input. Part II stresses internal structure and flow of program control and is intended to assist the researcher who wishes to revise or modify the code or add to its capabilities. PREREM is written for execution on a Digital Equipment Corporation PDP-10 System and much of the code will require revision before it can be run on other machines. The source program length is 950 lines (116 blocks) and computer core required for execution is 212 K bytes. The user must also have sufficient file space for metabolic and S-factor data sets. Further, 64 100 K byte blocks of computer storage space are required for the nuclear decay data file. Computer storage space must also be available for any output files produced during the PREREM execution. 9 refs., 8 tabs.

Fuel pin design algorithm for conceptual design studies

International Nuclear Information System (INIS)

Uselman, J.P.

1979-01-01

Two models are available which are currently verified by part of the requirements and which are adaptable as algorithms for the complete range. Fuel thermal performance is described by the HEDL SIEX model. Cladding damage and total deformation are determined by the GE GRO-II structural analysis code. A preliminary fuel pin performance model for analysis of (U, P/sub U/)O 2 pins in the COROPT core conceptual design system has been constructed by combining the key elements of SIEX and GRO-II. This memo describes the resulting pin performance model and its interfacing with COROPT system. Some exemplary results are presented

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-06-28

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the ²³⁵U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

Visual imagery and the user model applied to fuel handling at EBR-II

International Nuclear Information System (INIS)

Brown-VanHoozer, S.A.

1995-01-01

The material presented in this paper is based on two studies involving visual display designs and the user's perspective model of a system. The studies involved a methodology known as Neuro-Linguistic Programming (NLP), and its use in expanding design choices which included the ''comfort parameters'' and ''perspective reality'' of the user's model of the world. In developing visual displays for the EBR-II fuel handling system, the focus would be to incorporate the comfort parameters that overlap from each of the representation systems: visual, auditory and kinesthetic then incorporate the comfort parameters of the most prominent group of the population, and last, blend in the other two representational system comfort parameters. The focus of this informal study was to use the techniques of meta-modeling and synesthesia to develop a virtual environment that closely resembled the operator's perspective of the fuel handling system of Argonne's Experimental Breeder Reactor - II. An informal study was conducted using NLP as the behavioral model in a v reality (VR) setting

Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

1993-01-01

This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

Artificial Photosystem I and II: Highly Selective solar fuels and tandem photocatalysis

Science.gov (United States)

Ding, Yuchen; Castellanos, Ignacio; Cerkovnik, Logan; Nagpal, Prashant

2014-03-01

Artificial photosynthesis, or generation of solar fuels from CO2/H2O, can provide an important alternative for rising CO2 emission and renewable energy generation. In our recent work, composite photocatalysts (CPCs) made from widebandgap nanotubes and different QDs were used to mimic Photosystem II (PS680) and I (PS700), respectively. By tuning the redox potentials using the size, composition and energy band alignment of QDs, we demonstrate highly selective (>90%) and efficient production of ethane, ethanol and acetaldehyde as solar fuels with different wavelengths of light. We also show that this selectivity is a result of precise energy band alignments (using cationic/anionic doping of nanotubes, QD size etc.), confirmed using measurements of electronic density of states, and alignment of higher redox potentials with hot-carriers can also lead to hot-carrier photocatalysis. This wavelength-selective CPCs can have important implications for inexpensive production of solar fuels including alkanes, alcohols, aldehydes and hydrogen, and making tandem structures (red, green, blue) with three CPCs, allowing almost full visible spectrum (410 ~ 730nm) utilization with different fuels produced simultaneously.

The KNK II/1 fuel assembly NY-205: Compilation of the irradiation history and the fuel and fuel pin fabrication data of the INTERATOM data bank system BESEX

International Nuclear Information System (INIS)

Patzer, G.; Geier, F.

1988-01-01

The fuel assembly NY-205 has been irradiated during the first and the second core of KNK II with a total residence time of 832 equivalent full-power days. A maximum burnup of 175.000 MWd/tHM or 18.6 % was reached with a maximum steel damage of 66 dpa-NRT. For the cladding the materials 1.4970 and 1.4981 have been used in different metallurgical conditions, and for the Uranium/Plutonium mixed- oxide fuel the most important variants of the major fabrication parameters had been realized. The assembly will be brought to the Hot Cells of the KfK Karlsruhe for post-irradiation examination in February 1988, so that the knowledge of the fabrication data is of interest for the selection of fuel pins and for the evaluation of the examination results. Therefore this report compiles the fuel and fuel pin fabrication data from the INTERATOM data bank system BESEX and additionally, an overview of the irradiation history of the assembly is given [de

Calculus of Elementary Functions, Part II. Teacher's Commentary. Revised Edition.

Science.gov (United States)

Herriot, Sarah T.; And Others

This course is intended for students who have a thorough knowledge of college preparatory mathematics, including algebra, axiomatic geometry, trigonometry, and analytic geometry. This teacher's guide is for Part II of the course. It is designed to follow Part I of the text. The guide contains background information, suggested instructional…

Calculus of Elementary Functions, Part II. Student Text. Revised Edition.

Science.gov (United States)

Herriot, Sarah T.; And Others

This course is intended for students who have a thorough knowledge of college preparatory mathematics, including algebra, axiomatic geometry, trigonometry, and analytic geometry. This text, Part II, contains material designed to follow Part I. Chapters included in this text are: (6) Derivatives of Exponential and Related Functions; (7) Area and…

76 FR 67287 - Alternative Fuel Transportation Program; Alternative Fueled Vehicle Credit Program (Subpart F...

Science.gov (United States)

2011-10-31

... additional credits for the use of biodiesel in blends of 20 percent biodiesel or greater and have provided an... discussion in Part II.A), the original program based upon AFV acquisitions and biodiesel use became known as... example, B20 (a 20 percent blend of biodiesel with 80 percent petroleum diesel) is not an alternative fuel...

Converting Eucalyptus biomass into ethanol: Financial and sensitivity analysis in a co-current dilute acid process. Part II

International Nuclear Information System (INIS)

Gonzalez, R.; Treasure, T.; Phillips, R.; Jameel, H.; Saloni, D.; Wright, J.; Abt, R.

2011-01-01

The technical and financial performance of high yield Eucalyptus biomass in a co-current dilute acid pretreatment followed by enzymatic hydrolysis process was simulated using WinGEMS registered and Excel registered . Average ethanol yield per dry Mg of Eucalyptus biomass was approximately 347.6 L of ethanol (with average carbohydrate content in the biomass around 66.1%) at a cost of 0.49 L -1 of ethanol, cash cost of ∝0.46 L -1 and CAPEX of 1.03 L -1 of ethanol. The main cost drivers are: biomass, enzyme, tax, fuel (gasoline), depreciation and labor. Profitability of the process is very sensitive to biomass cost, carbohydrate content (%) in biomass and enzyme cost. Biomass delivered cost was simulated and financially evaluated in Part I; here in Part II the conversion of this raw material into cellulosic ethanol using the dilute acid process is evaluated. (author)

FRM-II project status and safety of its compact fuel element

International Nuclear Information System (INIS)

Nuding, M.; Rottmann, M.; Axmann, A.; Boening, K.

2000-01-01

The construction of the new research reactor FRM-II is close to completion and the nuclear start-up is scheduled to begin in January 2001. This contribution provides an overview on the concept of the facility and the safety features of the reactor. It also describes some of the tests performed during the licensing procedure of the compact fuel element and their results. At the end a short status report is given. (author)

FRM-II project status and safety of its compact fuel element

Energy Technology Data Exchange (ETDEWEB)

Nuding, M.; Rottmann, M.; Axmann, A.; Boening, K. [Technical University of Munich, D-85747 Garching (Germany)

2000-07-01

The construction of the new research reactor FRM-II is close to completion and the nuclear start-up is scheduled to begin in January 2001. This contribution provides an overview on the concept of the facility and the safety features of the reactor. It also describes some of the tests performed during the licensing procedure of the compact fuel element and their results. At the end a short status report is given. (author)

HANARO fuel irradiation test (II): revision

Energy Technology Data Exchange (ETDEWEB)

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

2001-04-01

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

Nursing Care of Patients Undergoing Chemotherapy Desensitization: Part II.

Science.gov (United States)

Jakel, Patricia; Carsten, Cynthia; Carino, Arvie; Braskett, Melinda

2016-04-01

Chemotherapy desensitization protocols are safe, but labor-intensive, processes that allow patients with cancer to receive medications even if they initially experienced severe hypersensitivity reactions. Part I of this column discussed the pathophysiology of hypersensitivity reactions and described the development of desensitization protocols in oncology settings. Part II incorporates the experiences of an academic medical center and provides a practical guide for the nursing care of patients undergoing chemotherapy desensitization.
.

40 CFR Appendix II to Part 1054 - Duty Cycles for Laboratory Testing

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Duty Cycles for Laboratory Testing II.... 1054, App. II Appendix II to Part 1054—Duty Cycles for Laboratory Testing (a) Test handheld engines with the following steady-state duty cycle: G3 mode No. Engine speed a Torque(percent) b Weighting...

Hydroprocessing and premium II refinery: a new refining philosophy for an era of clean fuels

Energy Technology Data Exchange (ETDEWEB)

Delgaudio, Caio Veiga Penna; Pinotti, Rafael [Petroleo Brasileiro S.A. (PETROBRAS), Rio de Janeiro, RJ (Brazil)

2012-07-01

This paper discusses a brief history of Brazilian's emission and fuel specifications, since the appearance of PROCONVE until the late stages of the program for vehicles powered by gasoline and diesel. The development of the Brazilian refining is analyzed taking into account the emission and specification evolutions, and it can be perceived that the system's complexity increases while new constraints are imposed by the regulator. This aspect is even more apparent when the detailed scheme of the Premium II refinery and its main unit, the catalytic hydrocracker (HCC, which has not yet been part of PETROBRAS' refining park and will appear in three of the four new refineries of the company) is described. The new projects represent the culmination of the intensive use of energy and raw material for obtaining the products with the new specifications. There is a price for this development, both in investments and increased operating costs due to greater complexity of the system. To adapt to the era of clean fuels, refiners will present a series of challenges that will lead them to seek for more efficient processes and operational excellence (and ongoing efforts to reduce their emissions) in order to ensure positive refining margins. (author)

Part 5. Fuel cycle options

International Nuclear Information System (INIS)

Lineberry, M.J.; McFarlane, H.F.; Amundson, P.I.; Goin, R.W.; Webster, D.S.

1980-01-01

The results of the FBR fuel cycle study that supported US contributions to the INFCE are presented. Fuel cycle technology is reviewed from both generic and historical standpoints. Technology requirements are developed within the framework of three deployment scenarios: the reference international, the secured area, and the integral cycle. Reprocessing, fabrication, waste handling, transportation, and safeguards are discussed for each deployment scenario. Fuel cycle modifications designed to increase proliferation defenses are described and assessed for effectiveness and technology feasibility. The present status of fuel cycle technology is reviewed and key issues that require resolution are identified

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

International Nuclear Information System (INIS)

Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

1977-01-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

1977-04-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Uranium accountability for ATR fuel fabrication: Part II. A computer simulation

International Nuclear Information System (INIS)

Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.

1977-08-01

A stochastic computer model has been designed to simulate the material control system used during the production of fuel plates for the Advanced Test Reactor. Great care has been taken to see that this model follows the manufacturing and measuring processes used. The model is designed so that manufacturing process and measurement parameters are fed in as input; hence, changes in the manufacturing process and measurement procedures are easily simulated. Individual operations in the plant are described by program subroutines. By varying the calling sequence of these subroutines, variations in the manufacturing process may be simulated. By using this model values for MUF and LEMUF may be calculated for predetermined plant operating conditions. Furthermore the effect on MUF and LEMUF produced by changing plant operating procedures and measurement techniques may also be examined. A sample calculation simulating one inventory period of the plant's operation is included

Market Analysis and Consumer Impacts Source Document. Part III. Consumer Behavior and Attitudes Toward Fuel Efficient Vehicles

Science.gov (United States)

1980-12-01

This source document on motor vehicle market analysis and consumer impacts consists of three parts. Part III consists of studies and reviews on: consumer awareness of fuel efficiency issues; consumer acceptance of fuel efficient vehicles; car size ch...

Probabilistic risk assessment on maritime spent nuclear fuel transportation (Part II: Ship collision probability)

International Nuclear Information System (INIS)

Christian, Robby; Kang, Hyun Gook

2017-01-01

This paper proposes a methodology to assess and reduce risks of maritime spent nuclear fuel transportation with a probabilistic approach. Event trees detailing the progression of collisions leading to transport casks’ damage were constructed. Parallel and crossing collision probabilities were formulated based on the Poisson distribution. Automatic Identification System (AIS) data were processed with the Hough Transform algorithm to estimate possible intersections between the shipment route and the marine traffic. Monte Carlo simulations were done to compute collision probabilities and impact energies at each intersection. Possible safety improvement measures through a proper selection of operational transport parameters were investigated. These parameters include shipment routes, ship's cruise velocity, number of transport casks carried in a shipment, the casks’ stowage configuration and loading order on board the ship. A shipment case study is presented. Waters with high collision probabilities were identified. Effective range of cruising velocity to reduce collision risks were discovered. The number of casks in a shipment and their stowage method which gave low cask damage frequencies were obtained. The proposed methodology was successful in quantifying ship collision and cask damage frequency. It was effective in assisting decision making processes to minimize risks in maritime spent nuclear fuel transportation. - Highlights: • Proposes a probabilistic framework on the safety of spent nuclear fuel transportation by sea. • Developed a marine traffic simulation model using Generalized Hough Transform (GHT) algorithm. • A transportation case study on South Korean waters is presented. • Single-vessel risk reduction method is outlined by optimizing transport parameters.

Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

International Nuclear Information System (INIS)

Ravnik, M.

1988-11-01

The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

Model of automatic fuel management for the Atucha II nuclear central with the PUMA IV code

International Nuclear Information System (INIS)

Marconi G, J.F.; Tarazaga, A.E.; Romero, L.D.

2007-01-01

The Atucha II central is a heavy water power station and natural uranium. For this reason and due to the first floor reactivity excess that have this type of reactors, it is necessary to carry out a continuous fuel management and with the central in power (for the case of Atucha II every 0.7 days approximately). To maintain in operation these centrals and to achieve a good fuels economy, different types of negotiate of fuels that include areas and roads where the fuels displace inside the core are proved; it is necessary to prove the great majority of these managements in long periods in order to corroborate the behavior of the power station and the burnt of extraction of the fuel elements. To carry out this work it is of great help that a program implements the approaches to continue in each replacement, using the roads and areas of each administration type to prove, and this way to obtain as results the one regulations execution in the time and the average burnt of extraction of the fuel elements, being fundamental this last data for the operator company of the power station. To carry out the previous work it is necessary that a physicist with experience in fuel management proves each one of the possible managements, even those that quickly can be discarded if its don't fulfill with the regulatory standards or its possess an average extraction burnt too much low. For this it is of fundamental help that with an automatic model the different administrations are proven and lastly the physicist analyzes the more important cases. The pattern in question not only allows to program different types of roads and areas of fuel management, but rather it also foresees the possibility to disable some of the approaches. (Author)

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

Science.gov (United States)

2010-10-01

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

Visual imagery and the user model applied to fuel handling at EBR-II

Energy Technology Data Exchange (ETDEWEB)

Brown-VanHoozer, S.A.

1995-06-01

The material presented in this paper is based on two studies involving visual display designs and the user`s perspective model of a system. The studies involved a methodology known as Neuro-Linguistic Programming (NLP), and its use in expanding design choices which included the ``comfort parameters`` and ``perspective reality`` of the user`s model of the world. In developing visual displays for the EBR-II fuel handling system, the focus would be to incorporate the comfort parameters that overlap from each of the representation systems: visual, auditory and kinesthetic then incorporate the comfort parameters of the most prominent group of the population, and last, blend in the other two representational system comfort parameters. The focus of this informal study was to use the techniques of meta-modeling and synesthesia to develop a virtual environment that closely resembled the operator`s perspective of the fuel handling system of Argonne`s Experimental Breeder Reactor - II. An informal study was conducted using NLP as the behavioral model in a v reality (VR) setting.

Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

Energy Technology Data Exchange (ETDEWEB)

Durkee, Jr., Joe W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-06-19

A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial ²³⁵U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for ¹³⁴Cs/¹³⁷Cs ¹³⁴Cs/¹⁵⁴Eu, and ¹⁵⁴Eu/¹³⁷Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial ²³⁵U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the

Current activities on improving storage conditions of the research reactor RA spent fuel - Part II

International Nuclear Information System (INIS)

Matausek, M.V.; Kopecni, M.; Vukadin, Z.; Plecas, I.; Pavlovic, R.; Sotic, O.; Marinkovic, N.

1998-01-01

To minimize further corrosion and preserve integrity of aluminum barrels and the stainless steel channel-type containers that were found to contain leaking spent fuel, actions to improve conditions in the existing spent fuel storage pool at the RA research reactor were initiated. Technology was elaborated and equipment was produced and applied for removal of sludge and other debris from the bottom of the pool, filtration of the pool water, sludge conditioning in cement matrix and disposal at the low and medium waste repository at VINCA site. More sophisticated operations are to be performed together with foreign experts. Safety measures and precautions were determined. Subcriticality was proved under normal and/or possible abnormal conditions. (author)

40 CFR Appendix II to Part 1039 - Steady-State Duty Cycles

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Steady-State Duty Cycles II Appendix... Appendix II to Part 1039—Steady-State Duty Cycles (a) The following duty cycles apply for constant-speed engines: (1) The following duty cycle applies for discrete-mode testing: D2 mode number Engine speed...

Coupled 3D neutronic and thermohydraulic calculations for a compact fuel element with disperse UMo fuel at FRM II

International Nuclear Information System (INIS)

Breitkreutz, H.; Roehrmoser, A.; Petry, W.

2010-01-01

The newly developed X 2 program system is intended to be used for high-detail 3D calculations on compact research reactor cores. Using this system, the efforts to calculate scenarios for a new fuel element for FRM II using disperse UMo (8wt% Mo, 50% enrichment) are continued. By now, a radial symmetric core model with averaged built-in components for the D 2 O tank is used. Two different scenarios are compared: The minimum fuel density of 7.5 g U/cm 3 and 8.0 g U/cm 3 with 60 days cycle length. In addition, two 'flux loss compensating' scenarios based on 8.0 g U/cm 3 with 10% higher power/longer reactor cycles are regarded. (author)

AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XI, PART I--MAINTAINING THE FUEL SYSTEM (PART I), CUMMINS DIESEL ENGINES, PART II--UNIT REPLACEMENT (ENGINE).

Science.gov (United States)

Human Engineering Inst., Cleveland, OH.

THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF DIFFERENCES BETWEEN TWO AND FOUR CYCLE ENGINES, THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM, AND THE PROCEDURES FOR DIESEL ENGINE REMOVAL. TOPICS ARE (1) REVIEW OF TWO CYCLE AND FOUR CYCLE CONCEPT, (2) SOME BASIC CHARACTERISTICS OF FOUR CYCLE ENGINES,…

First international 26Al interlaboratory comparison - Part II

International Nuclear Information System (INIS)

Merchel, Silke; Bremser, Wolfram

2005-01-01

After finishing Part I of the first international 26 Al interlaboratory comparison with accelerator mass spectrometry (AMS) laboratories [S. Merchel, W. Bremser, Nucl. Instr. and Meth. B 223-224 (2004) 393], the evaluation of Part II with radionuclide counting laboratories took place. The evaluation of the results of the seven participating laboratories on four meteorite samples shows a good overall agreement between laboratories, i.e. it does not reveal any statistically significant differences if results are compared sample-by-sample. However, certain interlaboratory bias is observed with a more detailed statistical analysis including some multivariate approaches

Review of the KBS II plan for handling and final storage of unreprocessed spent nuclear fuel

International Nuclear Information System (INIS)

1980-01-01

The Swedish utilities programme for disposal of spent nuclear fuel elements (KBS II) is summarized. Comments and criticism to the programme are given by experts from several foreign or international institutions. (L.E.)

Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

International Nuclear Information System (INIS)

Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S.

2013-01-01

The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

Programming Models for Three-Dimensional Hydrodynamics on the CM-5 (Part II)

International Nuclear Information System (INIS)

Amala, P.A.K.; Rodrigue, G.H.

1994-01-01

This is a two-part presentation of a timing study on the Thinking Machines CORP. CM-5 computer. Part II is given in this study and represents domain-decomposition and message-passing models. Part I described computational problems using a SIMD model and connection machine FORTRAN (CMF)

Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel; Evaluacion Analitica de 4 Metodos de Determinacion de PAHs medianteHPLC en un Fuel de Tipo II

Energy Technology Data Exchange (ETDEWEB)

Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

2009-05-21

An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

Scientific issues in fuel behaviour

International Nuclear Information System (INIS)

1995-01-01

The current limits on discharge burnup in today's nuclear power stations have proven the fuel to be very reliable in its performance, with a negligibly small rate of failure. However, for reasons of economy, there are moves to increase the fuel enrichment in order to extend both the cycle time and the discharge burnup. But, longer periods of irradiation cause increased microstructural changes in the fuel and cladding, implying a larger degradation of physical and mechanical properties. This degradation may well limit the plant life, hence the NSC concluded that it is of importance to develop a predictive capability of fuel behaviour at extended burnup. This can only be achieved through an improved understanding of the basic underlying phenomena of fuel behaviour. The Task Force on Scientific Issues Related to Fuel Behaviour of the NEA Nuclear Science Committee has identified the most important scientific issues on the subject and has assigned priorities. Modelling aspects are listed in Appendix A and discussed in Part II. In addition, quality assurance process for performing and evaluating new integral experiments is considered of special importance. Main activities on fuel behaviour modelling, as carried out in OECD Member countries and international organisations, are listed in Part III. The aim is to identify common interests, to establish current coverage of selected issues, and to avoid any duplication of efforts between international agencies. (author). refs., figs., tabs

Continuum Thermodynamics - Part II: Applications and Examples

Science.gov (United States)

Albers, Bettina; Wilmanski, Krzysztof

The intention by writing Part II of the book on continuum thermodynamics was the deepening of some issues covered in Part I as well as a development of certain skills in dealing with practical problems of oscopic processes. However, the main motivation for this part is the presentation of main facets of thermodynamics which appear when interdisciplinary problems are considered. There are many monographs on the subjects of solid mechanics and thermomechanics, on fluid mechanics and on coupled fields but most of them cover only special problems in great details which are characteristic for the chosen field. It is rather seldom that relations between these fields are discussed. This concerns, for instance, large deformations of the skeleton of porous materials with diffusion (e.g. lungs), couplings of deformable particles with the fluid motion in suspensions, couplings of adsorption processes and chemical reactions in immiscible mixtures with diffusion, various multi-component aspects of the motion, e.g. of avalanches, such as segregation processes, etc...

Steady- and transient-state analysis of fully ceramic microencapsulated fuel with randomly dispersed tristructural isotropic particles via two-temperature homogenized model-II: Applications by coupling with COREDAX

International Nuclear Information System (INIS)

Lee, Yoon Hee; Cho, Bum Hee; Cho, Nam Zin

2016-01-01

In Part I of this paper, the two-temperature homogenized model for the fully ceramic microencapsulated fuel, in which tristructural isotropic particles are randomly dispersed in a fine lattice stochastic structure, was discussed. In this model, the fuel-kernel and silicon carbide matrix temperatures are distinguished. Moreover, the obtained temperature profiles are more realistic than those obtained using other models. Using the temperature-dependent thermal conductivities of uranium nitride and the silicon carbide matrix, temperature-dependent homogenized parameters were obtained. In Part II of the paper, coupled with the COREDAX code, a reactor core loaded by fully ceramic microencapsulated fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure is analyzed via a two-temperature homogenized model at steady and transient states. The results are compared with those from harmonic- and volumetric-average thermal conductivity models; i.e., we compare keff eigenvalues, power distributions, and temperature profiles in the hottest single channel at a steady state. At transient states, we compare total power, average energy deposition, and maximum temperatures in the hottest single channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized model for Doppler temperature feedback lead to significant differences

Benchmark matrix and guide: Part II.

Science.gov (United States)

1991-01-01

In the last issue of the Journal of Quality Assurance (September/October 1991, Volume 13, Number 5, pp. 14-19), the benchmark matrix developed by Headquarters Air Force Logistics Command was published. Five horizontal levels on the matrix delineate progress in TQM: business as usual, initiation, implementation, expansion, and integration. The six vertical categories that are critical to the success of TQM are leadership, structure, training, recognition, process improvement, and customer focus. In this issue, "Benchmark Matrix and Guide: Part II" will show specifically how to apply the categories of leadership, structure, and training to the benchmark matrix progress levels. At the intersection of each category and level, specific behavior objectives are listed with supporting behaviors and guidelines. Some categories will have objectives that are relatively easy to accomplish, allowing quick progress from one level to the next. Other categories will take considerable time and effort to complete. In the next issue, Part III of this series will focus on recognition, process improvement, and customer focus.

Eutectic penetration times in irradiated EBR-II driver fuel elements

International Nuclear Information System (INIS)

Betten, P.R.; Bottcher, J.H.; Seidel, B.R.

1983-01-01

The experimental test procedure employed the use of a high-temperature furnace which heated pre-irradiated elements to temperature and maintained the environment until element-cladding breach occurred. Pre-irradiated elements of the Mark-II design were first encapsulated in a close-fitting sealed tube that was instrumented with a pressure transducer at the top of the tube and a thermocouple at the element's top-of-fuel axial location. The volume of the capsule was evacuated in order to better identify the pressure pulse which would occur on breach and to minimize contaminants. Next, a three-zone fast-recovery furnace was heated and an axial temperature profile, similar to that experienced in the EBR-II core, was established. The encapsulated element was then quickly inserted into the furnace and remained there until clad breach occurred. The element was then removed from the furnace immediately. Visual and metallurgical examination of the rupture site was done later. A total of seven elements were tested in the above manner

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Conceptual design of ICF reactor SENRI, Part II. Advances in design and pellet gain scaling

International Nuclear Information System (INIS)

Ido, S.; Mima, K.; Nakai, S.; Tsuji, R.; Yamanaka, C.

1984-01-01

This chapter reviews the recent design studies on reactor concepts with magnetically guided lithium flow, SENRI-I, SENRI-IA and SENRI-II. The routes from the present status to power reactors and an advanced fuel pellet concept is also discussed. Topics covered include pellet design, magnetohydrodynamic design of liquid lithium flow; reactor cavity concepts with magnetically guided lithium flow, a thermo-hydraulic analysis, a tritium recovery system; and an advanced fuel pellet concept for an inertial confinement fusion (ICF) reactor without a tritium breeding blanket. An advanced fuel pellet for an ICF reactor without a T breeder was studied in the model calculations, which showed sufficiently high values of pellet gain. Includes a table and 8 diagrams

Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning

International Nuclear Information System (INIS)

Eberle, C.S.; Dean, E.M.; Angelo, P.L.

1995-01-01

A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations

12 CFR Appendix II to Part 27 - Information for Government Monitoring Purposes

Science.gov (United States)

2010-01-01

... II Appendix II to Part 27 Banks and Banking COMPTROLLER OF THE CURRENCY, DEPARTMENT OF THE TREASURY... Monitoring Purposes The following language is approved by the Comptroller of the Currency and will satisfy... used separately. This information may also be provided orally by the applicant. The following...

31 CFR Appendix II to Part 13 - Form of Bill for Reimbursement

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 1 2010-07-01 2010-07-01 false Form of Bill for Reimbursement II Appendix II to Part 13 Money and Finance: Treasury Office of the Secretary of the Treasury PROCEDURES FOR... title) of ______ (Country) to participate in the work of ______ (International Organization) or...

Control console conceptual design for sheet type fuels of Triga Mark-II reactor

International Nuclear Information System (INIS)

Eko Priyono; Kurnia Wibowo; Anang Susanto

2016-01-01

The control console conceptual design for sheet type fuel of TRIGA Mark-II reactor has been made. The control console conceptual design was made with refer study result of instrument and control system which is used in BATAN'S reactor i.e TRIGA-2000 Bandung, TRIGA Yogyakarta and MPR-30 Serpong. The control console conceptual design was made by using AutoCad software. The control console conceptual design reactor for sheet type fuel of TRIGA Mark-II reactor consist of 5 segments that is 3 segments for placing the computer monitors, 1 segment for placing bargraph displays and recorders and 1 segment for placing panel meters. There are the door on front and back position at each segment for enter and out devices in the console. The control console conceptual design is also equipped by the table along in front of console for placing reactor panel control and for writing, 3 drawers for 3 keyboards. The dimension of console will refer control room size and the components will be placed on console which will be detailed in detail design if this conceptual design has been approved. (author)

Appendices. Part II

International Nuclear Information System (INIS)

1983-11-01

Information on LMFBR research activities is presented concerning reactor core kinetics and heat transfer characteristics; heat transfer and hydraulic reactions to transient conditions; fuel assembly power distribution and temperature gradients; and heat transfer and hydraulics associated with the recirculation experiment

10 CFR Appendix II to Part 1050 - DOE Form 3735.3-Foreign Travel Statement

Science.gov (United States)

2010-01-01

... is official agency business. Spouses and dependents may accept such travel and expenses only when... 10 Energy 4 2010-01-01 2010-01-01 false DOE Form 3735.3-Foreign Travel Statement II Appendix II to.... II Appendix II to Part 1050—DOE Form 3735.3—Foreign Travel Statement EC01OC91.041 Statement...

Part II: Oxidative Thermal Aging of Pd/Al2O3 and Pd/CexOy-ZrO2 in Automotive Three Way Catalysts: The Effects of Fuel Shutoff and Attempted Fuel Rich Regeneration

Directory of Open Access Journals (Sweden)

Qinghe Zheng

2015-10-01

Full Text Available The Pd component in the automotive three way catalyst (TWC experiences deactivation during fuel shutoff, a process employed by automobile companies for enhancing fuel economy when the vehicle is coasting downhill. The process exposes the TWC to a severe oxidative aging environment with the flow of hot (800 °C–1050 °C air. Simulated fuel shutoff aging at 1050 °C leads to Pd metal sintering, the main cause of irreversible deactivation of 3% Pd/Al2O3 and 3% Pd/CexOy-ZrO2 (CZO as model catalysts. The effect on the Rh component was presented in our companion paper Part I. Moderate support sintering and Pd-CexOy interactions were also experienced upon aging, but had a minimal effect on the catalyst activity losses. Cooling in air, following aging, was not able to reverse the metallic Pd sintering by re-dispersing to PdO. Unlike the aged Rh-TWCs (Part I, reduction via in situ steam reforming (SR of exhaust HCs was not effective in reversing the deactivation of aged Pd/Al2O3, but did show a slight recovery of the Pd activity when CZO was the carrier. The Pd+/Pd0 and Ce3+/Ce4+ couples in Pd/CZO are reported to promote the catalytic SR by improving the redox efficiency during the regeneration, while no such promoting effect was observed for Pd/Al2O3. A suggestion is made for improving the catalyst performance.

Fuel element failure detection experiments, evaluation of the experiments at KNK II/1 (Intermediate Report)

CERN Document Server

Bruetsch, D

1983-01-01

In the frame of the fuel element failure detection experiments at KNK II with its first core the measurement devices of INTERATOM were taken into operation in August 1981 and were in operation almost continuously. Since the start-up until the end of the first KNK II core operation plugs with different fuel test areas were inserted in order to test the efficiency of the different measuring devices. The experimental results determined during this test phase and the gained experiences are described in this report and valuated. All three measuring techniques (Xenon adsorption line XAS, gas-chromatograph GC and precipitator PIT) could fulfil the expectations concerning their susceptibility. For XAS and GC the nuclide specific sensitivities as determined during the preliminary tests could be confirmed. For PIT the influences of different parameters on the signal yield could be determined. The sensitivity of the device could not be measured due to a missing reference measuring point.

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P.O.1236909. Final report

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-12-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P. O. 1236909. Final report

Energy Technology Data Exchange (ETDEWEB)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-12-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design.

PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

International Nuclear Information System (INIS)

Anand, A.K.; Anantharaman, K.; Sarda, V.

1978-01-01

The Indian power programme envisages the use of PHWRs, which use collapsible clad UO 2 fuel elements. A computer code, PECITIS-II, developed for the analysis of this type of fuel is described in detail. The sheath strain and fission gas pressure are evaluated by this method. The pellet clad gap conductance is calculated by Ross and Solute model. The pellet thermal expansion is calculated by assuming a two zone model, i.e. a plastic core surrounded by an elastic cracked annulus. (author)

AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XXII, I--MAINTAINING THE FUEL SYSTEM (PART I)--CUMMINS DIESEL ENGINE, II--UNDERSTANDING THE DIFFERENTIAL.

Science.gov (United States)

Minnesota State Dept. of Education, St. Paul. Div. of Vocational and Technical Education.

THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE FUNCTION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM AND DIFFERENTIAL DRIVE UNITS USED IN DIESEL POWERED VEHICLES. TOPICS ARE (1) FUEL SYSTEM COMPARISONS, (2) FUEL SYSTEM SUPPLY COMPONENTS, (3) FUEL SUPPLY SECTION MAINTENANCE, (4) FUNCTION OF THE DIFFERENTIAL,…

AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XIII, I--MAINTAINING THE FUEL SYSTEM (PART III), CUMMINS DIESEL ENGINES, II--RADIATOR SHUTTER SYSTEM.

Science.gov (United States)

Human Engineering Inst., Cleveland, OH.

THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE CONSTRUCTION, OPERATION, AND MAINTENANCE OF THE DIESEL ENGINE FUEL AND RADIATOR SHUTTER SYSTEMS. TOPICS ARE (1) MORE ABOUT THE CUMMINS FUEL SYSTEM, (2) CALIBRATING THE PT FUEL PUMP, (3) CALIBRATING THE FUEL INJECTORS, (4) UNDERSTANDING THE SHUTTER SYSTEM, (5) THE…

Safety aspects of the IFR pyroprocess fuel cycle

International Nuclear Information System (INIS)

Forrester, R.J.; Lineberry, M.J.; Charak, I.; Tessier, J.H.; Solbrig, C.W.; Gabor, J.D.

1989-01-01

This paper addresses the important safety considerations related to the unique Integral Fast Reactor (IFR) fuel cycle technology, the pyroprocess. Argonne has been developing the IFR since 1984. It is a liquid metal cooled reactor, with a unique metal alloy fuel, and it utilizes a radically new fuel cycle. An existing facility, the Hot Fuel Examination Facility-South (HFEF/S) is being modified and equipped to provide a complete demonstration of the fuel cycle. This paper will concentrate on safety aspects of the future HFEF/S operation, slated to begin late next year. HFEF/S is part of Argonne's complex of reactor test facilities located on the Idaho National Engineering Laboratory. HFEF/S was originally put into operation in 1964 as the EBR-II Fuel Cycle Facility (FCF) (Stevenson, 1987). From 1964--69 FCF operated to demonstrate an earlier and incomplete form of today's pyroprocess, recycling some 400 fuel assemblies back to EBR-II. The FCF mission was then changed to one of an irradiated fuels and materials examination facility, hence the name change to HFEF/S. The modifications consist of activities to bring the facility into conformance with today's much more stringent safety standards, and, of course, providing the new process equipment. The pyroprocess and the modifications themselves are described more fully elsewhere (Lineberry, 1987; Chang, 1987). 18 refs., 5 figs., 2 tabs

Calculation of DND-signals in case of fuel pin failures in KNK II with the computer code FICTION III

International Nuclear Information System (INIS)

Schmuck, I.

1990-11-01

In KNK II two delayed neutron detectors are installed for quick detection of fuel subassembly cladding failures. They record the release of the precursors of the emitters of delayed neutrons into the sodium. The computer code FICTION III calculates the expected delayed neutron signals for certain fuel pin failures, where the user has to set the boundary conditions interactively. In view of FICTION II the advancement of FICTION III consists of the following items: application of the data sets of 105 isotopes, distinction of thermal and fast neutron induced fission, partitioning of the sodium flow into two circuits, consideration of the specific fission rates in 10 fuel pin sections, elaboration of the user's interaction possibilities for input/ output. The capability of FICTION III is shown by means of two applications (UNi-test pin on position 100 and the third KNK fuel subassembly cladding failure). Object of further evaluations will be among other things the analysis of increased delayed neutron signals in regard to the fault location and dimension

Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

2000-08-01

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect.

Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

International Nuclear Information System (INIS)

Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

2000-08-01

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect

SASSYS validation with the EBR-II shutdown heat removal tests

International Nuclear Information System (INIS)

Herzog, J.P.

1989-01-01

SASSYS is a coupled neutronic and thermal hydraulic code developed for the analysis of transients in liquid metal cooled reactors (LMRs). The code is especially suited for evaluating of normal reactor transients -- protected (design basis) and unprotected (anticipated transient without scram) transients. Because SASSYS is heavily used in support of the IFR concept and of innovative LMR designs, such as PRISM, a strong validation base for the code must exist. Part of the validation process for SASSYS is analysis of experiments performed on operating reactors, such as the metal fueled Experimental Breeder Reactor -- II (EBR-II). During the course of a series of historic whole-plant experiments, EBR-II illustrated key safety features of metal fueled LMRs. These experiments, the Shutdown Heat Removal Tests (SHRT), culminated in unprotected loss of flow and loss of heat sink transients from full power and flow. Analysis of these and earlier SHRT experiments constitutes a vital part of SASSYS validation, because it facilitates scrutiny of specific SASSYS models and of integrated code capability. 12 refs., 11 figs

Analysis of reactivity worths of highly-burnt PWR fuel samples measured in LWR-PROTEUS Phase II

Energy Technology Data Exchange (ETDEWEB)

Grimm, Peter; Murphy, Michael F.; Jatuff, Fabian; Seiler, Rudolf [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland)

2008-07-01

The reactivity loss of PWR fuel with burnup has been determined experimentally by inserting fresh and highly-burnt fuel samples in a PWR test lattice in the framework of the LWR-PROTEUS Phase II programme. Seven UO{sub 2} samples irradiated in a Swiss PWR plant with burnups ranging from approx40 to approx120 MWd/kg and four MOX samples with burnups up to approx70 MWd/kg were oscillated in a test region constituted of actual PWR UO{sub 2} fuel rods in the centre of the PROTEUS zero-power experimental facility. The measurements were analyzed using the CASMO-4E fuel assembly code and a cross section library based on the ENDF/B-VI evaluation. The results show close proximity between calculated and measured reactivity effects and no trend for a deterioration of the quality of the prediction at high burnup. The analysis thus demonstrates the high accuracy of the calculation of the reactivity of highly-burnt fuel. (authors)

A Survey of Optometry Graduates to Determine Practice Patterns: Part II: Licensure and Practice Establishment Experiences.

Science.gov (United States)

Bleimann, Robert L.; Smith, Lee W.

1985-01-01

A summary of Part II of a two-volume study of optometry graduates conducted by the Association of Schools and Colleges of Optometry is presented. Part II includes the analysis of the graduates' licensure and practice establishment experiences. (MLW)

Optimization of binary breeder reactor IV - Conception of mixed fuel at central part of the core

International Nuclear Information System (INIS)

Dias, A.F.; Ishiguro, Y.

1986-04-01

Neutronic characteristics of some LMFBRs are analized for a fueling mode that is different from those reported previously. In an inner part of the core both 233 U/ 232 Th and Pu/U assemblies are placed while the outer zone is fueled with Pu/U assemblies. Both oxide metal fuels and 232 Th and 238 U blankets are considered. (Author) [pt

Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

Persic, A.; Ravnik, M.; Zagar, T.

1998-01-01

Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

Energy Technology Data Exchange (ETDEWEB)

Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-07-28

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

BEHAVE: fire behavior prediction and fuel modeling system-BURN Subsystem, part 1

Science.gov (United States)

Patricia L. Andrews

1986-01-01

Describes BURN Subsystem, Part 1, the operational fire behavior prediction subsystem of the BEHAVE fire behavior prediction and fuel modeling system. The manual covers operation of the computer program, assumptions of the mathematical models used in the calculations, and application of the predictions.

Healing and relaxation in flows of helium II. Part II. First, second, and fourth sound

International Nuclear Information System (INIS)

Hills, R.N.; Roberts, P.H.

1978-01-01

In Part I of this series, a theory of helium II incorporating the effects of quantum healing and relaxation was developed. In this paper, the propagation of first, second, and fourth sound is discussed. Particular attention is paid to sound propagation in the vicinity of the lambda point where the effects of relaxation and quantum healing become important

Blade System Design Study. Part II, final project report (GEC).

Energy Technology Data Exchange (ETDEWEB)

Griffin, Dayton A. (DNV Global Energy Concepts Inc., Seattle, WA)

2009-05-01

As part of the U.S. Department of Energy's Low Wind Speed Turbine program, Global Energy Concepts LLC (GEC)1 has studied alternative composite materials for wind turbine blades in the multi-megawatt size range. This work in one of the Blade System Design Studies (BSDS) funded through Sandia National Laboratories. The BSDS program was conducted in two phases. In the Part I BSDS, GEC assessed candidate innovations in composite materials, manufacturing processes, and structural configurations. GEC also made recommendations for testing composite coupons, details, assemblies, and blade substructures to be carried out in the Part II study (BSDS-II). The BSDS-II contract period began in May 2003, and testing was initiated in June 2004. The current report summarizes the results from the BSDS-II test program. Composite materials evaluated include carbon fiber in both pre-impregnated and vacuum-assisted resin transfer molding (VARTM) forms. Initial thin-coupon static testing included a wide range of parameters, including variation in manufacturer, fiber tow size, fabric architecture, and resin type. A smaller set of these materials and process types was also evaluated in thin-coupon fatigue testing, and in ply-drop and ply-transition panels. The majority of materials used epoxy resin, with vinyl ester (VE) resin also used for selected cases. Late in the project, testing of unidirectional fiberglass was added to provide an updated baseline against which to evaluate the carbon material performance. Numerous unidirectional carbon fabrics were considered for evaluation with VARTM infusion. All but one fabric style considered suffered either from poor infusibility or waviness of fibers combined with poor compaction. The exception was a triaxial carbon-fiberglass fabric produced by SAERTEX. This fabric became the primary choice for infused articles throughout the test program. The generally positive results obtained in this program for the SAERTEX material have led to its

Comparison of thermal and radical effects of EGR gases on combustion process in dual fuel engines at part loads

International Nuclear Information System (INIS)

Pirouzpanah, V.; Khoshbakhti Saray, R.; Sohrabi, A.; Niaei, A.

2007-01-01

Dual fuel engines at part load inevitably suffer from lower thermal efficiency and higher emission of carbon monoxide and unburned fuel. This work is conducted to investigate the combustion characteristics of a dual fuel (Diesel-gas) engine at part loads using a single zone combustion model with detailed chemical kinetics for combustion of natural gas fuel. In this home made software, the presence of the pilot fuel is considered as a heat source that is deriving form two superposed Wiebe's combustion functions to account for its contribution to ignition of the gaseous fuel and the rest of the total released energy. The chemical kinetics mechanism consists of 112 reactions with 34 species. This combustion model is able to establish the development of the combustion process with time and the associated important operating parameters, such as pressure, temperature, heat release rate (HRR) and species concentration. Therefore, this work is an attempt to investigate the combustion phenomenon at part load and using exhaust gas recirculation (EGR) to improve the above mentioned problems. Also, the results of this work show that each of the different cases of EGR (thermal, chemical and radical cases) has an important role on the combustion process in dual fuel engines at part loads. It is found that all the different cases of EGR have positive effects on the performance and emission parameters of dual fuel engines at part loads despite the negative effect of some diluent gases in the chemical case, which moderates too much the positive effects of the thermal and radical cases of EGR. Predicted values show good agreement with corresponding experimental values over the whole range of engine operating conditions. Implications will be discussed in detail

Fight against fuel poverty. Levers, stakes and expectations of the fight against fuel poverty in housing

International Nuclear Information System (INIS)

Payen, Luc; Pamart, Isabelle; Lacroix, Olivier

2013-10-01

'Is in fuel poverty a person who feels in his particular housing difficulties have the necessary energy supply to the satisfaction of basic needs due to the inadequacy of resources or its habitat conditions'. The rising cost of energy commodities in the late 2000's, added to the poor thermal quality an important part of French homes, has led to the emergence of fuel poverty in the public debate. Legislative recognition of these situations with the law 'Grenelle II' (from which is extracted the definition above) marked a decisive step in the fight against this complex problem. Affecting nearly 5 million households in France, fuel poverty is a major challenge for societies wishing to successfully achieve their energy transition. In this new publication, ENEA reports on the main levers of the fight against fuel poverty, the obstacles encountered and the needs for new solutions

The Mid America Heart Institute: part II.

Science.gov (United States)

McCallister, Ben D; Steinhaus, David M

2003-01-01

The Mid America Heart Institute (MAHI) is one of the first and largest hospitals developed and designed specifically for cardiovascular care. The MAHI hybrid model, which is a partnership between the not-for-profit Saint Luke's Health System, an independent academic medical center, and a private practice physician group, has been extremely successful in providing high-quality patient care as well as developing strong educational and research programs. The Heart Institute has been the leader in providing cardiovascular care in the Kansas City region since its inception in 1975. Although challenges in the future are substantial, it is felt that the MAHI is in an excellent position to deal with the serious issues in health care because of the Heart Institute, its facility, organization, administration, dedicated medical and support staff, and its unique business model of physician management. In part I, the authors described the background and infrastructure of the Heart Institute. In part II, cardiovascular research and benefits of physician management are addressed.

Automotive fuels survey. Part 4. Innovations or illusions

International Nuclear Information System (INIS)

Troelstra, W.P.; Van Walwijk, M.; Bueckmann, M.

1999-01-01

Volumes 1 to 3 of the IEA/AFIS Automotive Fuels Survey, address the most well-known automotive fuels and fuel production routes. Less well-known fuels and energy sources that are not used in combustion engines, e.g. electricity, were excluded from these volumes. In this report fuel routes and fuels that have not been addressed in the first volumes will be analysed. In this report, each chapter starts with a short description of the fuel(route) and its status of development (e.g. if the idea has been abandoned or if the fuel is already sold at a fuel station). Then the different aspects of that fuel are described as far as the information is available. This is limited to information that can not be found in volumes one and two of the Automotive Fuels Survey. For example: for the diesel-water mixtures, the production of diesel is not be described. If comparisons are made, they are made either relative to an already described fuel(route) that is related (e.g. biogas will be compared with natural gas) or relative to diesel and gasoline as was done in volume 1 and 2 of the Automotive Fuels Survey. For some of the fuels, the relation with a fuel already covered in volume one and two is very strong. For these fuels more information can be found in the chapters on the related fuel in the other volumes of the Automotive Fuels Survey. The following fuels are covered in this report: biodiesel from used oil and fat, biodiesel and biogasoline from algae, diesel from hydrothermal upgrading, biogas, hythane, Fischer-Tropsch diesel, diesel-water blends, higher ethers, and electricity. 74 refs

Marketing in the E-Business World, Parts I & II | Smith | LBS ...

African Journals Online (AJOL)

Marketing in the E-Business World, Parts I & II. ... Open Access DOWNLOAD FULL TEXT ... of many of Americas largest companies gather at the Waldorf Astoria Hotel in New York City for the Conference Boards Annual Marketing Conference.

Evironmental assessment factors relating to reprocessing of spent nuclear fuel

International Nuclear Information System (INIS)

1978-05-01

This document is in two parts. Part I presents the criteria and evaluation factors, based primarily on US experience, which may be used to carry out an environmental assessment of spent fuel reprocessing. The concept of As Low as is Reasonably Achievable (ALARA) is introduced in limiting radiation exposure. The factors influencing both occupational and general public radiation exposure are reviewed. Part II provides information on occupational and general public radiation exposure in relation to reprocessing taken from various sources including UNSCEAR and GESMO. Some information is provided in relation to potential accidents at reprocessing or MOX fuel refabrication plants. The magnitude of the services, energy, land use and non-radiological effluents for the reference design of reprocessing plant are also presented

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

International Nuclear Information System (INIS)

Murray, A.M.

1999-01-01

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

40 CFR Appendix II to Part 1045 - Duty Cycles for Propulsion Marine Engines

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Duty Cycles for Propulsion Marine... Pt. 1045, App. II Appendix II to Part 1045—Duty Cycles for Propulsion Marine Engines (a) The following duty cycle applies for discrete-mode testing: E4 Mode No. Enginespeed 1 Torque(percent) 2...

Electrometallurgical treatment of aluminum-matrix fuels

International Nuclear Information System (INIS)

Willit, J.L.; Gay, E.C.; Miller, W.E.; McPheeters, C.C.; Laidler, J.J.

1996-01-01

The electrometallurgical treatment process described in this paper builds on our experience in treating spent fuel from the Experimental Breeder Reactor (EBR-II). The work is also to some degree, a spin-off from applying electrometallurgical treatment to spent fuel from the Hanford single pass reactors (SPRs) and fuel and flush salt from the Molten Salt Reactor Experiment (MSRE) in treating EBR-II fuel, we recover the actinides from a uranium-zirconium fuel by electrorefining the uranium out of the chopped fuel. With SPR fuel, uranium is electrorefined out of the aluminum cladding. Both of these processes are conducted in a LiCl-KCl molten-salt electrolyte. In the case of the MSRE, which used a fluoride salt-based fuel, uranium in this salt is recovered through a series of electrochemical reductions. Recovering high-purity uranium from an aluminum-matrix fuel is more challenging than treating SPR or EBR-II fuel because the aluminum- matrix fuel is typically -90% (volume basis) aluminum

Methods of humidity determination Part II: Determination of material humidity

OpenAIRE

Rübner, Katrin; Balköse, Devrim; Robens, E.

2008-01-01

Part II covers the most common methods of measuring the humidity of solid material. State of water near solid surfaces, gravimetric measurement of material humidity, measurement of water sorption isotherms, chemical methods for determination of water content, measurement of material humidity via the gas phase, standardisation, cosmonautical observations are reviewed.

Nonlinear observer designs for fuel cell power systems

Science.gov (United States)

Gorgun, Haluk

A fuel cell is an electrochemical device that combines hydrogen and oxygen, with the aid of electro-catalysts, to produce electricity. A fuel cell consists of a negatively charged anode, a positively charged cathode and an electrolyte, which transports protons or ions. A low temperature fuel cell has an electrical potential of about 0.7 Volt when generating a current density of 300--500 mA/cm2. Practical fuel cell power systems will require a combination of several cells in series (a stack) to satisfy the voltage requirements of specific applications. Fuel cells are suitable for a potentially wide variety of applications, from stationary power generation in the range of hundreds of megawatts to portable electronics in the range of a couple of watts. Efficient operation of a fuel cell system requires advanced feedback control designs. Reliable measurements from the system are necessary to implement such designs. However, most of the commercially available sensors do not operate properly in the reformate and humidified gas streams in fuel cell systems. Sensors working varying degrees of success are too big and costly, and sensors that are potentially low cost are not reliable or do not have the required life time [28]. Observer designs would eliminate sensor needs for measurements, and make feedback control implementable. Since the fuel cell system dynamics are highly nonlinear, observer design is not an easy task. In this study we aim to develop nonlinear observer design methods applicable to fuel cell systems. In part I of the thesis we design an observer to estimate the hydrogen partial pressure in the anode channel. We treat inlet partial pressure as an unknown slowly varying parameter and develop an adaptive observer that employs a nonlinear voltage injection term. However in this design Fuel Processing System (FPS) dynamics are not modelled, and their effect on the anode dynamics are treated as plant uncertainty. In part II of the thesis we study the FPS

Implementation of the II. Stage decommissioning of A1 NPP

International Nuclear Information System (INIS)

Ficher, T.

2015-01-01

Presentation is focused on the implementation of the II. stage decommissioning of A1 NPP. Introductory part focuses on brief characteristics of the power plant with a history of operation, basic technical parameters and actions that were made after operation. The next section describes the basic schedule for decommissioning, structure of management and implementation of the II. stage decommissioning of the A1 NPP and objectives of the individual stages. The last and largest part of the presentation is devoted to detailed description of the II. stage decommissioning of the A1 NPP, its individual tasks and verbal and visual description of the activities that were performed. Presented is decommissioning of the technology and construction of external objects NPP A1 including storage tanks for liquid RAW, next are presented activities carried out in the Main Production Unit - decommissioning of non-operating technologies in various places/rooms, management of waste arising from these activities, treatment of case of A1 long-term spent fuel storage and long-term spent fuel storage. The subsequent section is devoted to the management and handling of contaminated soil, concrete and construction waste, including management of VLLW. (authors)

Feasibility study on commercialization of fast breeder reactor cycle systems interim report of phase II. Technical study report for nuclear fuel cycle systems

International Nuclear Information System (INIS)

Sato, Koji; Amamoto, Ippei; Inoue, Akira

2004-06-01

As a part of the feasibility study on commercialization of fast breeder reactor cycle systems, the plant concept concerning the fuel cycle systems (combination of the reprocessing and the fuel fabrication) has been constructed to reduce their total cost by the introduction of various innovative techniques and to apply their utmost superior efficiency from such standpoints of a decrease in the environmental burden, better resource utilization and proliferation resistance improvement by the low decontamination transuranium element (TRU) recycle. This interim report of Phase II describes the results of an on-going study which will cover a five-year period. For oxide fuels, the system which combines the use of the advanced aqueous reprocessing using three main methods such as the crystallization method, the simplified solvent extraction method, and the extraction chromatography method for minor actinide (MA) recovery, as well as the simplified pelletizing fuel fabrication which rationalized a powder mixing process etc., has abundant current results and a high technical feasibility for the basic process. Though this system faces difficulties in the technical development of control technology of the extraction chromatography and the fabrication technology of low decontamination TRU fuel etc., its expected practical use is possible at an early stage. As for the super-critical direct extraction reprocessing, it is necessary to fulfill more basic data although further economical improvement of an advanced aqueous reprocessing is expected. The system which combines the advanced aqueous reprocessing and the gelation sphere packing fuel fabrication has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the fuel fabrication process. However, this system will shoulder additional cost for the reagent recovery process and the waste liquid treatment process due to need to dispose of a large bulk of process waste liquid. The system which

A cold demonstration of fuel consolidation. Part 1

International Nuclear Information System (INIS)

Matheson, J.E.

1989-01-01

Spent fuel consolidation is an option for increasing spent fuel storage capacities being considered by many utilities. The process of consolidating fuel involves separating the fuel rods from the structural frame which holds them in a square array. The rods are then repackaged into a tightly packed bundle which occupies about half the cross-sectional area of fuel assembly. Thus approximately twice as much fuel can be stored in the underwater racks at a spent fuel storage pool. There have been several demonstrations of fuel consolidation to date. The focus of this paper is the development and subsequent demonstration program of a shear/compactor

The year 2012 in the European Heart Journal-Cardiovascular Imaging. Part II.

Science.gov (United States)

Plein, Sven; Knuuti, Juhani; Edvardsen, Thor; Saraste, Antti; Piérard, Luc A; Maurer, Gerald; Lancellotti, Patrizio

2013-07-01

The part II of the best of the European Heart Journal - Cardiovascular Imaging in 2012 specifically focuses on studies of valvular heart diseases, heart failure, cardiomyopathies, and congenital heart diseases.

Pyroprocessing of oxidized sodium-bonded fast reactor fuel - An experimental study of treatment options for degraded EBR-II fuel

Energy Technology Data Exchange (ETDEWEB)

Hermann, S.D.; Gese, N.J. [Separations Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Wurth, L.A. [Zinc Air Inc., 5314-A US Hwy 2 West, Columbia Falls, MT 59912 (United States)

2013-07-01

An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electro-metallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li{sub 2}O at 650 C. degrees with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. In the absence of zirconium or sodium oxide, the electrolytic reduction of MnO showed nearly complete conversion to metal. The electrolytic reduction of a blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O showed substantial reduction of manganese, but only 8.5% of the zirconium was found in the metal phase. The electrolytic reduction of the same blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O - 6.2 wt% Na{sub 2}O showed substantial reduction of manganese, but zirconium reduction was even less at 2.4%. This study concluded that ZrO{sub 2} cannot be substantially reduced to metal in an electrolytic reduction system with LiCl - 1 wt% Li{sub 2}O at 650 C. degrees due to the perceived preferential formation of lithium zirconate. This study also identified a possible interference that sodium oxide may have on the same system by introducing a parasitic and cyclic reaction of dissolved sodium metal between oxidation at the anode and reduction at the cathode. When applied to oxidized sodium-bonded EBR-II fuel (e.g., U-10Zr), the prescribed electrolytic reduction system would not be expected to substantially reduce zirconium oxide, and the accumulation of sodium in the electrolyte could interfere with the reduction of uranium oxide, or at least render it less efficient.

Storage of Spent Nuclear Fuel. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

On the Efficiency of Connection Charges---Part II: Integration of Distributed Energy Resources

OpenAIRE

Munoz-Alvarez, Daniel; Garcia-Franco, Juan F.; Tong, Lang

2017-01-01

This two-part paper addresses the design of retail electricity tariffs for distribution systems with distributed energy resources (DERs). Part I presents a framework to optimize an ex-ante two-part tariff for a regulated monopolistic retailer who faces stochastic wholesale prices on the one hand and stochastic demand on the other. In Part II, the integration of DERs is addressed by analyzing their endogenous effect on the optimal two-part tariff and the induced welfare gains. Two DER integrat...

Fuels for homogeneous charge compression ignition (HCCI) engines. Automotive fuels survey. Part 6

Energy Technology Data Exchange (ETDEWEB)

Van Walwijk, M.

2001-01-01

Homogeneous charge compression ignition (HCCI) is a third mode of operation for internal combustion engines, beside spark ignition and conventional compression ignition. This report concentrates on the requirements that HCCI operation puts on fuels for these engines. For readers with limited time available, this summary describes the main findings. Policy makers that need some more background information may turn directly to chapter 7, 'Fuels for HCCI engines'. The rest of this report can be considered as a reference guide for more detailed information. The driving force to investigate HCCI engines is the potential of low emissions and simultaneously high energy efficiency. HCCI is gaining attention the last few years. However, HCCI engines are still in the research phase. After many experiments with prototype engines, people have now started working on computer simulations of the combustion process, to obtain a fundamental understanding of HCCI combustion and to steer future engine developments. In HCCI engines, an air/fuel mixture is prepared before it enters the combustion chamber. The homogeneous mixture is in the combustion chamber compressed to auto-ignition. Unlike in conventional engines, combustion starts at many different locations simultaneously and the speed of combustion is very high, so there is no flame front. Lean air/fuel mixtures (excess air) are used to control combustion speed. Because of the excess air, combustion temperature is relatively low, resulting in low NOx emissions. When the fuel is vaporised to a truly homogeneous mixture, complete combustion results in low particulate emissions. The most important advantages of HCCI engines are: - Emissions of NOx and particulates are very low. - Energy efficiency is high. It is comparable to diesel engines. - Many different fuels (one at a time) can be used in the HCCI concept. There are also some hurdles to overcome: - Controlling combustion is difficult, it complicates engine design

Development of metal fuel and study of construction materials (I-IV), Part II; Razvoj metalnog goriva i ispitivanje konstrukcionih materijala (I-VI deo); II deo

Energy Technology Data Exchange (ETDEWEB)

Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-11-15

The studies were devoted to problems related to application of metal uranium as fuel in heavy water reactors. Influence of thermal treatment on material texture and recrystallization of cast uranium was investigated. Structural changes of uranium alloys with molybdenum and niobium were tested during different heat treatments. A review of the possibilities for using metal uranium fuel in heavy water reactors is included.

The optimization of spent fuel assembly storage racks in nuclear power plants

International Nuclear Information System (INIS)

Wang Yan

2005-01-01

This paper gives an evaluation of the spent fuel assembly storage racks in the nuclear power plants at home and abroad, focusing on the characteristics of the high density storage racks and the aseismatic design. It mainly discusses structures and characteristics of the spent fuel assembly storage racks in the Qinshan nuclear power phase II project. Concluding the crucial technical difficulties of the high density spent fuel assembly storage racks: the neutron-absorbing materials, the structural aseismatic design technology and the security analysis technology, this paper firstly generalizes several important neutron-absorbing materials, then introduces the evolution of the aseismatic design of the spent fuel assembly storage racks . In the last part, it describes the advanced aseismatic analysis technology in the Qinshan nuclear power phase II project. Through calculation and analysis for such storage racks, the author concludes several main factors that could have an influence on the aseismatic performance and thus gives the key points and methods for designing the optimal racks and provides some references for the design of advanced spent fuel assembly storage racks in the future. (authors)

76 FR 18066 - Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program

Science.gov (United States)

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...-generating foreign producers and importers of renewable fuels for which RINs have been generated by the...

Final disposal of spent fuels and high activity waste: status and trends in the world. Part 2

International Nuclear Information System (INIS)

Herscovich de Pahissa, Marta

2008-01-01

The proper management of spent fuel arising from nuclear power production is a key issue for the sustainable development of nuclear energy. Some countries have adopted reprocessing of spent fuel and part of them has continued to develop and improve closed fuel cycle technologies; some other countries have adopted a direct final disposal. The objective in this article is to provide an update on the latest development in the world related with the geological disposal of spent nuclear fuel and high level wastes. (author) [es

The investigation of fast reactor fuel pin start up behaviour in the irradiation experiment DUELL II

International Nuclear Information System (INIS)

Freund, D.; Geithoff, D.

1988-04-01

The irradiation experiments DUELL-II within the SNR-300 operational Transient Experimental Program deal with the investigation of fresh mixed oxide fuel behaviour at start-up. The irradiation has been carried out in the HFR Petten in four so-called DUELL capsules with two fuel pin samples each. The fuel pins with a total length of 453 mm contained a fuel column of 150 mm length, consisting of high dense (U,Pu)O 2-x fuel with an initial porosity of 4%, a Pu-content of 20.9%, and an O/Me ratio of 1.96. The fuel pellet diameter was 6.37 mm, the outer diameter of the SS cladding, material No. 1.4970, was 7.6 mm. The irradiation included four phases, consisting of preconditioning at 85% nominal power (corresponds to 550 W/cm), a following increase to full power, and two following full power periods of 1 and 10 days, respectively. Post irradiation examination showed incomplete fuel restructuring in the first capsules with central void diameters of 800 μm in the hot plane, complete restructuring in the last capsule, leading to central voids of approximately 1 mm diameter. The residual gaps between fuel and clad varied between 25 and 44 μm. The clad inner surface did not show any corrosion attack. The analysis of fuel restructuring has been carried out with the computer code SATURN-S showing good agreement with the PIE results. The analysis led to a series of model improvements, especially for crack volume and relocation modelling. (orig./GL) [de

Alternate-Fueled Combustor-Sector Performance—Part A: Combustor Performance and Part B: Combustor Emissions

OpenAIRE

Shouse, D. T.; Neuroth, C.; Hendricks, R. C.; Lynch, A.; Frayne, C. W.; Stutrud, J. S.; Corporan, E.; Hankins, Capt. T.

2012-01-01

Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F or ASTM D 7566 standards, respectively, and are classified as “drop-in’’ fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are acceptable. Adherence to alternate fuels and fuel blends requires “smart fueling systems’’ or advanced fuel-flexible systems, including combustors and engines, without significant sacrifice in performance or emissions requirements...

Hydrogen Fuel Cells: Part of the Solution

Science.gov (United States)

Busby, Joe R.; Altork, Linh Nguyen

2010-01-01

With the decreasing availability of oil and the perpetual dependence on foreign-controlled resources, many people around the world are beginning to insist on alternative fuel sources. Hydrogen fuel cell technology is one answer to this demand. Although modern fuel cell technology has existed for over a century, the technology is only now becoming…

Subseabed disposal program annual report, January-December 1979. Volume II. Appendices (principal investigator progress reports). Part 2 of 2

International Nuclear Information System (INIS)

Talbert, D.M.

1981-04-01

Volume II of the sixth annual report describing the progress and evaluating the status of the Subseabed Disposal Program contains the appendices referred to in Volume II, Summary and Status. Because of the length of Volume II, it has been split into two parts for publication purposes. Part 1 contains Appendices A-O; Part 2 contains Appendices P-FF. Separate abstracts have been prepared for each appendix for inclusion in the Energy Data Base

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

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.

1994-10-01

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

Irradiation of Parts of the X-Gen Nuclear Fuel Assembly made by KNF in HANARO

International Nuclear Information System (INIS)

Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Lee, S. H.; Eom, K. B.

2008-01-01

An instrumented capsule has been developed at HANARO (High flux Advanced Neutron Application ReactOr) for the neutron irradiation tests of materials. The capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. As a preliminary test, some specimens made of the parts of a nuclear fuel assembly were inserted in the 05M-07U instrumented capsule and successfully irradiated at HANARO. Based on the results and experience, a new irradiation capsule of 07M-13N was designed, fabricated, and irradiated at HANARO for the evaluation of the neutron irradiation properties of the parts of the X-Gen nuclear fuel assembly made by KNF (Korea Nuclear Fuel). Specimens such as bucking and spring test specimens of spacer grid, microstructure and tensile test specimens of welded parts, tensile, irradiation growth and spring test specimens made of HANA tube, Zirlo, Zircaloy-4 and Inconel-718 were placed in the capsule. The capsule was loaded into the CT test hole of HANARO of a 30MW thermal output and the specimens were irradiated at 295 - 460 .deg. C up to a fast neutron fluence of 1.2x10 21 (n/cm 2 ) (E>1.0MeV)

Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

International Nuclear Information System (INIS)

Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

1986-05-01

In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

Subseabed disposal program annual report, January-December 1980. Volume II. Appendices (principal investigator progress reports). Part 1

International Nuclear Information System (INIS)

Hinga, K.R.

1981-07-01

Volume II of the sixth annual report describing the progress and evaluating the status of the Subseabed Disposal Program contains the appendices referred to in Volume I, Summary and Status. Because of the length of Volume II, it has been split into two parts for publication purposes. Part 1 contains Appendices A-Q; Part 2 contains Appendices R-MM. Separate abstracts have been prepared for each appendix for inclusion in the Energy Data Base

Subseabed disposal program annual report, January-December 1980. Volume II. Appendices (principal investigator progress reports). Part 1

Energy Technology Data Exchange (ETDEWEB)

Hinga, K.R. (ed.)

1981-07-01

Volume II of the sixth annual report describing the progress and evaluating the status of the Subseabed Disposal Program contains the appendices referred to in Volume I, Summary and Status. Because of the length of Volume II, it has been split into two parts for publication purposes. Part 1 contains Appendices A-Q; Part 2 contains Appendices R-MM. Separate abstracts have been prepared for each appendix for inclusion in the Energy Data Base.

Fuel cells science and engineering. Materials, processes, systems and technology. Vol. 1

Energy Technology Data Exchange (ETDEWEB)

Stolten, Detlef; Emonts, Bernd (eds.) [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF), Brennstoffzellen (IEF-3)

2012-07-01

The first volume is divided in four parts and 22 chapters. It is structured as follows: PART I: Technology. Chapter 1: Technical Advancement of Fuel-Cell Research and Development (Dr. Bernd Emonts, Ludger Blum, Thomas Grube, Werner Lehnert, Juergen Mergel, Martin Mueller and Ralf Peters); 2: Single-Chamber Fuel Cells (Teko W. Napporn and Melanie Kuhn); 3: Technology and Applications of Molten Carbonate Fuel Cells (Barbara Bosio, Elisabetta Arato and Paolo Greppi); 4: Alkaline Fuel Cells (Erich Guelzow); 5: Micro Fuel Cells (Ulf Groos and Dietmar Gerteisen); 6: Principles and Technology of Microbial Fuel Cells (Jan B. A. Arends, Joachim Desloover, Sebastia Puig and Willy Verstraete); 7: Micro-Reactors for Fuel Processing (Gunther Kolb); 8: Regenerative Fuel Cells (Martin Mueller). PART II: Materials and Production Processes. Chapter 9: Advances in Solid Oxide Fuel Cell Development between 1995 and 2010 at Forschungszentrum Juelich GmbH, Germany (Vincent Haanappel); 10: Solid Oxide Fuel Cell Electrode Fabrication by Infiltration (Evren Gunen); 11: Sealing Technology for Solid Oxide Fuel Cells (K. Scott Weil); 12: Phosphoric Acid, an Electrolyte for Fuel Cells - Temperature and Composition Dependence of Vapor Pressure and Proton Conductivity (Carsten Korte); 13: Materials and Coatings for Metallic Bipolar Plates in Polymer Electrolyte Membrane Fuel Cells (Heli Wang and John A. Turner); 14: Nanostructured Materials for Fuel Cells (John F. Elter); 15: Catalysis in Low-Temperature Fuel Cells - An Overview (Sabine Schimpf and Michael Bron). PART III: Analytics and Diagnostics. Chapter 16: Impedance Spectroscopy for High-Temperature Fuel Cells (Ellen Ivers-Tiffee, Andre Leonide, Helge Schichlein, Volker Sonn and Andre Weber); 17: Post-Test Characterization of Solid Oxide Fuel-Cell Stacks (Norbert H. Menzler and Peter Batfalsky); 18: In Situ Imaging at Large-Scale Facilities (Christian Toetzke, Ingo Manke and Werner Lehnert); 19: Analytics of Physical Properties of Low

Starting a hospital-based home health agency: Part II--Key success factors.

Science.gov (United States)

Montgomery, P

1993-09-01

In Part II of a three-part series, the financial, technological and legislative issues of a hospital-based home health-agency are discussed. Beginning a home healthcare service requires intensive research to answer key environmental and operational questions--need, competition, financial projections, initial start-up costs and the impact of delayed depreciation. Assessments involving technology, staffing, legislative and regulatory issues can help project service volume, productivity and cost-control.

TREAT hodoscope interpretation. II. Fuel-state identification

International Nuclear Information System (INIS)

Wu, R.M.; Omberg, R.P.; Albrecht, R.W.

1982-01-01

By using the autoregressive-integrated-moving-average (ARIMA) process, the onset of fuel disposal and the restructured fuel states of a TREAT test can be unambiguously identified. The results of the ARIMA analyses on the TREAT L7 hodoscope data show the most probable time of the restructuring began at 14.038 seconds, and four restructured fuel states are required to interpret adequately the L7 hodoscope data

Association Between National Board Dental Examination Part II Scores and Comprehensive Examinations at Harvard School of Dental Medicine.

Science.gov (United States)

Lee, Min Kyeong; Allareddy, Veerasathpurush; Howell, T Howard; Karimbux, Nadeem Y

2011-01-01

Harvard School of Dental Medicine (HSDM) uses a hybrid problem-based approach to teaching in the predoctoral program. The objective structured clinical examination (OSCE) is a formative examination designed to assess the performance of students in the problem-based learning (PBL) curriculum. At HSDM three comprehensive examinations with OSCE components are administered during the third and fourth years of clinical training. The National Board Dental Examination (NBDE) Part II is taken in the final year of the predoctoral program. This study examines the association between the NBDE Part II and the comprehensive exams held at HSDM. Predoctoral students from the HSDM classes of 2005 and 2006 were included in this study. The outcome variable of interest was the scores obtained by students in the NBDE Part II, and the main independent variable of interest was the performance of students in the comprehensive exams (honors, pass, make-up exam to pass). The Mann-Whitney U-test was used to examine the association between the grades obtained in the each of the three comprehensive exams and the NBDE Part II scores. Multivariable linear regression analysis was also used to examine the association between the NBDE Part II scores and the comprehensive exam grades. The effect of potential confounding factors including age, sex, and race/ethnicity was adjusted. The results suggest that students who performed well in the comprehensive exams performed better on the NBDE Part II, even after adjusting for confounding factors. Future studies will examine the long-term impact of PBL on postdoctoral plans and career choices.

Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

International Nuclear Information System (INIS)

Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

1976-02-01

An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

Safety analysis and optimization of the core fuel reloading for the Moroccan TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Nacir, B.; Boulaich, Y.; Chakir, E.; El Bardouni, T.; El Bakkari, B.; El Younoussi, C.

2014-01-01

Highlights: • Additional fresh fuel elements must be added to the reactor core. • TRIGA reactor could safely operate around 2 MW power with 12% fuel elements. • Thermal–hydraulic parameters were calculated and the safety margins are respected. • The 12% fuel elements will have no influence on the safety of the reactor. - Abstract: The Moroccan TRIGA MARK II reactor core is loaded with 8.5% in weight of uranium standard fuel elements. Additional fresh fuel elements must periodically be added to the core in order to remedy the observed low power and to return to the initial reactivity excess at the End Of Cycle. 12%-uranium fuel elements are available to relatively improve the short fuel lifetime associated with standard TRIGA elements. These elements have the same dimensions as standards elements, but with different uranium weight. The objective in this study is to demonstrate that the Moroccan TRIGA reactor could safely operate, around 2 MW power, with new configurations containing these 12% fuel elements. For this purpose, different safety related thermal–hydraulic parameters have been calculated in order to ensure that the safety margins are largely respected. Therefore, the PARET model for this TRIGA reactor that was previously developed and combined with the MCNP transport code in order to calculate the 3-D temperature distribution in the core and all the most important parameters like the axial distribution of DNBR (Departure from Nucleate Boiling Ratio) across the hottest channel. The most important conclusion is that the 12% fuel elements utilization will have no influence on the safety of the reactor while working around 2 MW power especially for configurations based on insertions in C and D-rings

Irradiation performance of metallic fuels

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Porter, D.L.; Batte, G.L.; Hofman, G.L.

1989-01-01

Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

Combustion chemistry and flame structure of furan group biofuels using molecular-beam mass spectrometry and gas chromatography - Part II: 2-Methylfuran.

Science.gov (United States)

Tran, Luc-Sy; Togbé, Casimir; Liu, Dong; Felsmann, Daniel; Oßwald, Patrick; Glaude, Pierre-Alexandre; Fournet, René; Sirjean, Baptiste; Battin-Leclerc, Frédérique; Kohse-Höinghaus, Katharina

2014-03-01

This is Part II of a series of three papers which jointly address the combustion chemistry of furan and its alkylated derivatives 2-methylfuran (MF) and 2,5-dimethylfuran (DMF) under premixed low-pressure flame conditions. Some of them are considered to be promising biofuels. With furan as a common basis studied in Part I of this series, the present paper addresses two laminar premixed low-pressure (20 and 40 mbar) flat argon-diluted (50%) flames of MF which were studied with electron-ionization molecular-beam mass spectrometry (EI-MBMS) and gas chromatography (GC) for equivalence ratios φ=1.0 and 1.7, identical conditions to those for the previously reported furan flames. Mole fractions of reactants, products as well as stable and reactive intermediates were measured as a function of the distance above the burner. Kinetic modeling was performed using a comprehensive reaction mechanism for all three fuels given in Part I and described in the three parts of this series. A comparison of the experimental results and the simulation shows reasonable agreement, as also seen for the furan flames in Part I before. This set of experiments is thus considered to be a valuable additional basis for the validation of the model. The main reaction pathways of MF consumption have been derived from reaction flow analyses, and differences to furan combustion chemistry under the same conditions are discussed.

77 FR 1319 - Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel Standards

Science.gov (United States)

2012-01-09

... fuels such as ethanol and biodiesel. Potentially regulated categories include: NAICS \\1\\ Examples of... they are produced as well as the cost associated with transporting these fuels to the U.S. Of the... II.E, we believe that the 1.0 billion gallon standard can indeed be met. Since biodiesel has an...

Intelligent control of HVAC systems. Part II: perceptron performance analysis

Directory of Open Access Journals (Sweden)

Ioan URSU

2013-09-01

Full Text Available This is the second part of a paper on intelligent type control of Heating, Ventilating, and Air-Conditioning (HVAC systems. The whole study proposes a unified approach in the design of intelligent control for such systems, to ensure high energy efficiency and air quality improving. In the first part of the study it is considered as benchmark system a single thermal space HVAC system, for which it is assigned a mathematical model of the controlled system and a mathematical model(algorithm of intelligent control synthesis. The conception of the intelligent control is of switching type, between a simple neural network, a perceptron, which aims to decrease (optimize a cost index,and a fuzzy logic component, having supervisory antisaturating role for neuro-control. Based on numerical simulations, this Part II focuses on the analysis of system operation in the presence only ofthe neural control component. Working of the entire neuro-fuzzy system will be reported in a third part of the study.

Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

International Nuclear Information System (INIS)

Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

1978-09-01

In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

International Working Group on Fast Reactors Thirteenth Annual Meeting. Summary Report. Part II

International Nuclear Information System (INIS)

1980-10-01

The Thirteenth Annual Meeting of the IAEA International Working Group on Fast Reactors was held at the IAEA Headquarters, Vienna, Austria from 9 to 11 April 1980. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programme in the field of LMFBRs and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

Oil from biomass corncob tar as a fuel

International Nuclear Information System (INIS)

Zhang, Hongmei; Wang, Jun

2007-01-01

In this study, biomass corncob tar oil (B-oil I and B-oil II) was extracted and its characteristics were measured. The characterization data show some similarities and differences among B-oil I, B-oil II and the Diesel: flash point. The densities and viscosities are higher than that of Diesel fuel. The solidifying point for B-oil I and B-oil II were lower than that of Diesel. The heating value of B-oil I and B-oil II were about 85.6% and 87.3% of that ordinary Diesel fuel (OD). The distillation temperatures of B-oil I and B-oil II were lower than that of Diesel fuel, with the 50% evaporation point being as much as 10 o C and 4 o C lower and the 90% evaporation point being 10 o C and 2 o C lower, respectively. These evaporation characteristics implied better cold starting and warm up properties of B-oil I and B-oil II than that of Diesel fuel. B-oil I and B-oil II were blended with Diesel in 10% and 20% by volume. Engine tests have been conducted with the aim of obtaining comparative measures of torque, thermal efficiency, specific fuel consumption and emissions such as CO, smoke density and NO to evaluate and compute the behavior of the Diesel engine running on the above mentioned fuels. The reduction in exhaust emissions, together with the increases in torque and thermal efficiency and the reduction in specific fuel consumption made the blends of B-oil I and B-oil II a suitable alternative fuel for Diesel and could help in controlling air pollution

Comparison of thermal, radical and chemical effects of EGR gases using availability analysis in dual-fuel engines at part loads

International Nuclear Information System (INIS)

Hosseinzadeh, A.; Khoshbakhti Saray, R.; Seyed Mahmoudi, S.M.

2010-01-01

Dual-fuel engines at part load inevitably suffer from lower thermal efficiency and higher emission of carbon monoxide and unburned fuel. A quasi-two-zone combustion model has been developed for studying the second-law analysis of a dual-fuel (diesel-gas) engine operating under part-load conditions. The model is composed of two divisions: a single-zone combustion model with chemical kinetics for combustion of natural gas fuel and a subsidiary zone for combustion of pilot fuel. In the latter zone, the pilot fuel is considered as a heat source derived from two superposed Wiebe's combustion functions to account for contribution of pilot fuel in ignition of gaseous fuel and the rest of the total released energy. This quasi-two-zone combustion model is able to establish the development of combustion process with time and associated important operating parameters, such as pressure, temperature, heat release rate (HRR) and species concentration. The present work is an attempt to investigate the combustion phenomenon from second-law point of view at part load and using exhaust gas recirculation (EGR) to improve the aforementioned problems. Therefore, the availability analysis is applied to the engine from inlet valve closing (IVC) until exhaust valve opening (EVO). Various availability components are identified and calculated separately with crank position. In this paper, the various availability components are identified and calculated separately with crank position. Then the different cases of EGR (chemical, radical and thermal cases) are applied to the availability analysis in dual-fuel engines at part loads. It is found that the chemical case of EGR has negative effect and in this case the unburned chemical availability is increased and the work availability decreases in comparison with baseline engine (without EGR). While the thermal and radical cases have positive effects on the availability terms especially on the unburned chemical availability and work availability

Design support document for the K Basins Vertical Fuel Handling Tools

International Nuclear Information System (INIS)

Bridges, A.E.

1995-01-01

The purpose of this document is to provide the design support information for the Vertical Fuel Handling Tools, developed for the removal of N Reactor fuel elements from their storage canisters in the K Basins storage pool and insertion into the Single Fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N Reactor fuel elements is part of the overall characterization effort. These new hand tools are required since previous fuel movement has involved grasping the fuel in a horizontal position. These tools are required to lift an element vertically from the storage canister. Additionally, a Mark II storage canister Lip Seal Protector was designed and fabricated for use during fuel retrieval. This device was required to prevent damage to the canister lip should a fuel element accidentally be dropped during its retrieval, using the handling tools. Supporting documentation for this device is included in this document

Nuclear Fuel Cycle System Analysis (II)

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Kwon, Eun Ha; Yoon, Ji Sup; Park, Seong Won

2007-04-15

As a nation develops strategies that provide nuclear energy while meeting its various objectives, it must begin with identification of a fuel cycle option that can be best suitable for the country. For such a purpose, this paper takes four different fuel cycle options that are likely adopted by the Korean government, considering the current status of nuclear power generation and the 2nd Comprehensive Nuclear Energy Promotion Plan (CNEPP) - Once-through Cycle, DUPIC Recycle, Thermal Reactor Recycle and GEN-IV Recycle. The paper then evaluates each option in terms of sustainability, environment-friendliness, proliferation-resistance, economics and technologies. Like all the policy decision, however, a nuclear fuel cycle option can not be superior in all aspects of sustainability, environment-friendliness, proliferation-resistance, economics, technologies and so on, which makes the comparison of the options extremely complicated. Taking this into consideration, the paper analyzes all the four fuel cycle options using the Multi-Attribute Utility Theory (MAUT) and the Analytic Hierarchy Process (AHP), methods of Multi-Attribute Decision Making (MADM), that support systematical evaluation of the cases with multi- goals or criteria and that such goals are incompatible with each other. The analysis shows that the GEN-IV Recycle appears to be most competitive.

Delivery systems for biopharmaceuticals. Part II: Liposomes, Micelles, Microemulsions and Dendrimers.

Science.gov (United States)

Silva, Ana C; Lopes, Carla M; Lobo, José M S; Amaral, Maria H

2015-01-01

Biopharmaceuticals are a generation of drugs that include peptides, proteins, nucleic acids and cell products. According to their particular molecular characteristics (e.g. high molecular size, susceptibility to enzymatic activity), these products present some limitations for administration and usually parenteral routes are the only option. To avoid these limitations, different colloidal carriers (e.g. liposomes, micelles, microemulsions and dendrimers) have been proposed to improve biopharmaceuticals delivery. Liposomes are promising drug delivery systems, despite some limitations have been reported (e.g. in vivo failure, poor long-term stability and low transfection efficiency), and only a limited number of formulations have reached the market. Micelles and microemulsions require more studies to exclude some of the observed drawbacks and guarantee their potential for use in clinic. According to their peculiar structures, dendrimers have been showing good results for nucleic acids delivery and a great development of these systems during next years is expected. This is the Part II of two review articles, which provides the state of the art of biopharmaceuticals delivery systems. Part II deals with liposomes, micelles, microemulsions and dendrimers.

Fuel and nuclear fuel cycle

International Nuclear Information System (INIS)

Prunier, C.

1998-01-01

The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

Signs of revision in Don Quixote, Part II

Directory of Open Access Journals (Sweden)

Gonzalo Pontón

2016-11-01

Full Text Available This article provides new evidences in favour of the hypothesis that Cervantes, after finishing Don Quixote, Part II, partially revised the original, introducing some significant changes and additions, mainly in the last chapters. The analysis of some narrative inconsistencies, that cannot be interpreted as mere mistakes but as significant textual traces, reveals a process of re-elaboration –a process that affects at least four sections of the novel. Most of the evidence gathered here suggests that this revision is closely linked to Avellaneda’s continuation, in the sense that Cervantes tried to challenge the apocriphal Quixote making last-time interventions in his own text.

Study of brushless fuel pump (improvement of pump and motor parts). 2nd Report. Blushless dendo fuel pump no kento. 2

Energy Technology Data Exchange (ETDEWEB)

Mine, K; Takada, S; Tatematsu, M; Takeuchi, H [Aisan Industry Co. Ltd., Aichi (Japan)

1992-10-01

A methanol use electrically driven fuel pump was developed as reported in the present report. Mixed fuel of gasoline with alcohol can be handled by a brushless fuel pump which was proposed and improved as reported. The flow rate performance was heightened to 25g/sec by heightening in output power of motor, while the high temperature performance was 17% heightened against the conventional ratio of lowering in flow rate by heightening in vapor jet capacity. Against the corrosiveness of methanol, an in-tank type was applied to the pump, and all its electrically conductive and other mechanical parts were made to be both anti-corrosive and anti-abrasive. It is structurally of a two-stage series turbine type of non-volume form. A sensor method was applied to the motor by confining the miniaturized control circuit of brushless motor in the motor so that the transistor is controlled against the heightening in temperature. The motor is a three-phase half-wave driving motor. Also developed was a fuel supply system which is useful for the mixed fuel covering a range of 100% methanol through 100% gasoline. The present pump is dimensionally interchangeable with the conventional gasoline use one. Its operational life is more than 10000 hours. 3 refs., 17 figs., 1 tab.

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part II; Redovisning av saekerhet efter foerslutning av slutfoervaret foer anvaent kaernbraensle. Huvudrapport fraan projekt SR-Site. Del II

Energy Technology Data Exchange (ETDEWEB)

2011-07-01

The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

Simulation and operation of the EBR-II automatic control rod drive system

International Nuclear Information System (INIS)

Lehto, W.K.; Larson, H.A.; Dean, E.M.; Christensen, L.J.

1985-01-01

An automatic control rod drive system (ACRDS) installed at EBR-II produces shaped power transients from 40% to full reactor power at a linear ramp rate of 4 MWt/s. A digital computer and modified control-rod-drive provides this capability. Simulation and analysis of ACRDS experiments establish the safety envelope for reactor transient operation. Tailored transients are required as part of USDOE Operational Reliability Testing program for prototypic fast reactor fuel cladding breach behavior studies. After initial EBR-II driver fuel testing and system checkout, test subassemblies were subjected to both slow and fast transients. In addition, the ACRDS is used for steady-state operation and will be qualified to control power ascent from initial critical to full power

Simulation and operation of the EBR-II automatic control rod drive system

International Nuclear Information System (INIS)

Lehto, W.K.; Larson, H.A.; Dean, E.M.; Christensen, L.J.

1985-01-01

An automatic control rod drive system (ACRDS) installed at EBR-II produces shaped power transients from 40% to full reactor power at a linear ramp rate of 4 MWt/s. A digital computer and modified control-rod-drive provides this capability. Simulation and analysis of ACRDS experiments establish the safety envelope for reactor transient operation. Tailored transients are required as part of USDOE Operational Reliability Testing program for prototypic fast reactor fuel cladding breach behavior studies. After initial EBR-II driver fuel testing and system checkout, test subassemblies were subjected to both slow and fast transients. In additions, the ACRDS is used for steady-state operation and will be qualified to control power ascent from initial critical to full power

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

The EBR-II spent fuel treatment program

International Nuclear Information System (INIS)

Lineberry, M.J.; McFarlane, H.F.

1995-01-01

Argonne National Laboratory has refurbished and equipped an existing hot cell facility for demonstrating a high-temperature electrometallurgical process for treating spent nuclear fuel from the Experimental Breeder Reactor-11. Two waste forms will be produced and qualified for geologic disposal of the fission and activation products. Relatively pure uranium will be separated for storage. Following additional development, transuranium elements will be blended into one of the high-level waste streams. The spent fuel treatment program will help assess the viability of electrometallurgical technology as a spent fuel management option

Nursing as concrete philosophy, Part II: Engaging with reality.

Science.gov (United States)

Theodoridis, Kyriakos

2018-04-01

This is the second paper of an essay in two parts. The first paper (Part I) is a critical discussion of Mark Risjord's conception of nursing knowledge where I argued against the conception of nursing knowledge as a kind of nursing science. The aim of the present paper (Part II) is to explicate and substantiate the thesis of nursing as a kind of concrete philosophy. My strategy is to elaborate upon certain themes from Wittgenstein's Tractatus in order to canvass a general scheme of philosophy based on a distinction between reality and the world. This distinction will be employed in the appropriation of certain significant features of nursing and nursing knowledge. By elaborating on the contrast between the abstract and the concrete, I will suggest that nursing may be seen as a kind of concrete philosophy, being primarily concerned with reality (and secondarily with the world). This thesis, I will argue, implies that philosophy is the kind of theory that is essential to nursing (which is not so much a theory than a certain kind of activity). © 2017 John Wiley & Sons Ltd.

Dry Process Fuel Performance Evaluation

Energy Technology Data Exchange (ETDEWEB)

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

2005-04-15

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

Dry Process Fuel Performance Evaluation

International Nuclear Information System (INIS)

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

2005-04-01

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

Emission factors of air pollutants from CNG-gasoline bi-fuel vehicles: Part I. Black carbon.

Science.gov (United States)

Wang, Yang; Xing, Zhenyu; Xu, Hui; Du, Ke

2016-12-01

Compressed natural gas (CNG) is considered to be a "cleaner" fuel compared to other fossil fuels. Therefore, it is used as an alternative fuel in motor vehicles to reduce emissions of air pollutants in transportation. To quantify "how clean" burning CNG is compared to burning gasoline, quantification of pollutant emissions under the same driving conditions for motor vehicles with different fuels is needed. In this study, a fleet of bi-fuel vehicles was selected to measure the emissions of black carbon (BC), carbon monoxide (CO), hydrocarbon (HC) and nitrogen oxide (NO x ) for driving in CNG mode and gasoline mode respectively under the same set of constant speeds and accelerations. Comparison of emission factors (EFs) for the vehicles burning CNG and gasoline are discussed. This part of the paper series reports BC EFs for bi-fuel vehicles driving on the real road, which were measured using an in situ method. Our results show that burning CNG will lead to 54%-83% reduction in BC emissions per kilometer, depending on actual driving conditions. These comparisons show that CNG is a cleaner fuel than gasoline for motor vehicles in terms of BC emissions and provide a viable option for reducing BC emissions cause by transportation. Copyright © 2016 Elsevier B.V. All rights reserved.

Fuel gases

International Nuclear Information System (INIS)

Anon.

1996-01-01

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

Impedance-Source Networks for Electric Power Conversion Part II

DEFF Research Database (Denmark)

Siwakoti, Yam P.; Peng, Fang Zheng; Blaabjerg, Frede

2015-01-01

Impedance-source networks cover the entire spectrum of electric power conversion applications (dc-dc, dc-ac, ac-dc, ac-ac) controlled and modulated by different modulation strategies to generate the desired dc or ac voltage and current at the output. A comprehensive review of various impedance......-source-network-based power converters has been covered in a previous paper and main topologies were discussed from an application point of view. Now Part II provides a comprehensive review of the most popular control and modulation strategies for impedance-source network-based power converters/inverters. These methods...

TRIGA Mark II Ljubljana - spent fuel transportation

International Nuclear Information System (INIS)

Ravnik, M.; Dimic, V.

2008-01-01

The most important activity in 1999 was shipment of the spent fuel elements back to the United States for final disposal. This activity started already in 1998 with some governmental support. In July 1999 all spent fuel elements (219 pieces) from the TRIGA research reactor in Ljubljana were shipped back to the United Stated by the ship from the port Koper in Slovenia. At the same time shipment of the spent fuel from the research reactor in Pitesti, Romania, and the research reactor in Rome, Italy, was conducted. During the loading the radiation exposure to the workers was rather low. The loading and shipment of the spent nuclear fuel went very smoothly and according the accepted time table. During the last two years the TRIGA research reactor in Ljubljana has been in operation about 1100 hours per year and without any undesired shut-down. (authors)

Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

International Nuclear Information System (INIS)

Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

1999-08-01

The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

Life-cycle analysis of energy and greenhouse gas emissions of automotive fuels in India: Part 1 – Tank-to-Wheel analysis

International Nuclear Information System (INIS)

Gupta, S.; Patil, V.; Himabindu, M.; Ravikrishna, R.V.

2016-01-01

As part of a two-part life cycle efficiency and greenhouse gas emission analysis for various automotive fuels in the Indian context, this paper presents the first part, i.e., Tank-to-Wheel analysis of various fuel/powertrain configurations for a subcompact passenger car. The Tank-to-Wheel analysis was applied to 28 fuel/powertrain configurations using fuels such as gasoline, diesel, compressed natural gas, liquefied petroleum gas and hydrogen with various conventional and hybrid electric powertrains. The gasoline-equivalent fuel economy and carbon dioxide emission results for individual fuel/powertrain configuration are evaluated and compared. It is found that the split hybrid configuration is best among hybrids as it leads to fuel economy improvement and carbon dioxide emissions reduction by 20–40% over the Indian drive cycle. Further, the engine efficiency, engine on-off time and regenerative braking energy assessment is done to evaluate the causes for higher energy efficiency of hybrid electric vehicles. The hybridization increases average engine efficiency by 10–60% which includes 19–23% of energy recovered at wheel through regenerative braking over the drive cycle. Overall, the Tank-to-Wheel energy use and efficiency results are evaluated for all fuel/powertrain configurations which show Battery Electric Vehicle, fuel cell vehicles and diesel hybrids are near and long term energy efficient vehicle configurations. - Highlights: • Tank-to-Wheel energy use & CO_2 emissions for subcompact car on Indian driving cycle. • Gasoline, diesel, CNG, LPG, hydrogen and electric vehicles are evaluated in this study. • First comprehensive Tank-to-Wheel analysis for India on small passenger car platform. • Parallel, series and split hybrid electric vehicles with various fuels are analysed.

Proposed fuel cycle for the Integral Fast Reactor

International Nuclear Information System (INIS)

Burris, L.; Walters, L.C.

1985-01-01

One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor

Transient and steady-state analyses of an electrically heated Topaz-II Thermionic Fuel Element

International Nuclear Information System (INIS)

El-Genk, M.S.; Xue, H.

1992-01-01

Transient and steady-state analyses of electrically heated, Thermionic Fuel Elements (TFEs) for Topaz-II space power system are performed. The calculated emitter and collector temperatures, load electric power and conversion efficiency are in good agreement with reported data. In this paper the effects or Cs pressure, thermal power input, and load resistance on the steady-state performance of the TFE are also investigated. In addition, the thermal response of the ZrH moderator during a startup transient and following a change in the thermal power input is examined

The "Pseudocommando" mass murderer: part II, the language of revenge.

Science.gov (United States)

Knoll, James L

2010-01-01

In Part I of this article, research on pseudocommandos was reviewed, and the important role that revenge fantasies play in motivating such persons to commit mass murder-suicide was discussed. Before carrying out their mass shootings, pseudocommandos may communicate some final message to the public or news media. These communications are rich sources of data about their motives and psychopathology. In Part II of this article, forensic psycholinguistic analysis is applied to clarify the primary motivations, detect the presence of mental illness, and discern important individual differences in the final communications of two recent pseudocommandos: Seung-Hui Cho (Virginia Tech) and Jiverly Wong (Binghamton, NY). Although both men committed offenses that qualify them as pseudocommandos, their final communications reveal striking differences in their psychopathology.

CE and nanomaterials - Part II: Nanomaterials in CE.

Science.gov (United States)

Adam, Vojtech; Vaculovicova, Marketa

2017-10-01

The scope of this two-part review is to summarize publications dealing with CE and nanomaterials together. This topic can be viewed from two broad perspectives, and this article is trying to highlight these two approaches: (i) CE of nanomaterials, and (ii) nanomaterials in CE. The second part aims at summarization of publications dealing with application of nanomaterials for enhancement of CE performance either in terms of increasing the separation resolution or for improvement of the detection. To increase the resolution, nanomaterials are employed as either surface modification of the capillary wall forming open tubular column or as additives to the separation electrolyte resulting in a pseudostationary phase. Moreover, nanomaterials have proven to be very beneficial for increasing also the sensitivity of detection employed in CE or even they enable the detection (e.g., fluorescent tags of nonfluorescent molecules). © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Expert's statement on the research reactor Munich II (FRM-II); Gutachterliche Stellungnahme zum Forschungsreaktor Muenchen II (FRM-II)

Energy Technology Data Exchange (ETDEWEB)

Liebert, Wolfgang; Friess, Friederike; Gufler, Klaus; Arnold, Nikolaus [Univ. fuer Bodenkultur (BOKU), Wien (Austria). Inst. fuer Sicherheits- und Risikowissenschaften (ISR)

2017-12-15

The Expert's statement on the research reactor FRM-II covers the following issues: The situation in Germany with respect to HEU (highly enriched uranium) fuel elements, the proliferation problems related to HEU fuel and the generated high-level radioactive wastes, possible safety hazards of an interim storage of HEU containing wastes, for instance in the interim storage facility Ahaus, possible safety hazards of final disposal of HEU containing radioactive wastes, possibilities to avoid the use of HEU fuel in order to prevent further production of these wastes, requirement of processing spent HEU containing fuel elements for final disposal.

Sludge in the pulp and paper industry in Sweden, part II[Combustion of]; Slam fraan skogsindustrin, fas II

Energy Technology Data Exchange (ETDEWEB)

Gyllenhammar, Marianne; Herstad Svaerd, Solvie; Kjoerk, Anders; Larsson, Sara; Wennberg, Olle [S.E.P. Scandinavian Energy Project AB, Goeteborg (Sweden); Aamand, Lars-Erik [Chalmers Univ. of Technology, Goeteborg (Sweden); Eskilsson, David [Swedish National Testing and Research Inst., Boraas (Sweden)

2003-11-01

During part II of this research program combustible sludge from the pulp and paper industry has been studied in detail. 560,000 tonnes of sludge per year (calculated as dry sludge) are produced in Sweden. The energy potential in the produced sludge is about 2 TWh/year. Today 1 TWh/year is produced in the pulp and paper mill's own boilers. This means that additional energy can be utilized from this material. An objective of this program has been to decide whether or not there are sludge types which are favourable respectively difficult to combust. By mixing different sludge types, or other waste products, emissions and/or problems during combustion can be minimized. These possibilities have been studied thoroughly in this program. A lot of sludge samples have been studied in laboratory scale at SP and in full-scale at Chalmers 12 MW CFB boiler. As a complement to the practical tests S.E.P. has done research regarding different aspects of sludge as a fuel; for example handling of sludge and regional drying. The results of 40 sintering tests at SP showed that the sintering temperature during combustion of sludge in a fluidised bed, with silica sand as bed material, varied between <850 deg C and >1100 deg C. The evaluation showed that the alkali content in the ash had the largest influence on the sintering temperature. Other factors were less important. During the tests at Chalmers eleven different sludge samples have been combusted together with wood pellets. Initially there were problems with the feeding to the boiler for some of the sludge samples. When the fuel feeding problems were solved the combustion took place without any problems. When sludge is co-combusted together with a 'clean' base fuel such as wood pellets the sulphur-, nitrogen- and chloride contents in the sludge have a large impact on the emissions. The normal way to reduce sulphur dioxide but also hydrogen chloride is to add lime in different positions into and after the boiler. In

Management of super-grade plutonium in spent nuclear fuel

International Nuclear Information System (INIS)

McFarlane, H. F.; Benedict, R. W.

2000-01-01

This paper examines the security and safeguards implications of potential management options for DOE's sodium-bonded blanket fuel from the EBR-II and the Fermi-1 fast reactors. The EBR-II fuel appears to be unsuitable for the packaging alternative because of DOE's current safeguards requirements for plutonium. Emerging DOE requirements, National Academy of Sciences recommendations, draft waste acceptance requirements for Yucca Mountain and IAEA requirements for similar fuel also emphasize the importance of safeguards in spent fuel management. Electrometallurgical treatment would be acceptable for both fuel types. Meeting the known requirements for safeguards and security could potentially add more than $200M in cost to the packaging option for the EBR-II fuel

Water activities in Forsmark (Part II). The final disposal facility for spent fuel: water activities above ground

International Nuclear Information System (INIS)

Werner, Kent; Hamren, Ulrika; Collinder, Per; Ridderstolpe, Peter

2010-09-01

The construction of the repository for spent nuclear fuel in Forsmark is associated with a number of measures above ground that constitute water operations according to Chapter 11 in the Swedish Environmental Code. This report, which is an appendix to the Environmental Impact Assessment, describes these water operations, their effects and consequences, and planned measures

40 CFR Appendix III to Part 266 - Tier II Emission Rate Screening Limits for Free Chlorine and Hydrogen Chloride

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Tier II Emission Rate Screening Limits for Free Chlorine and Hydrogen Chloride III Appendix III to Part 266 Protection of Environment... to Part 266—Tier II Emission Rate Screening Limits for Free Chlorine and Hydrogen Chloride Terrain...

Modifications to HFEF/S for IFR fuel cycle demonstration

International Nuclear Information System (INIS)

Lineberry, M.J.; Phipps, R.D.; Forrester, R.J.; Carnes, M.D.; Rigg, R.H.

1988-01-01

Modifications have begun to the Hot Fuel Examination Facility-South (HFEF/S) in order to demonstrate the technology of the integral fast reactor (IFR) fuel cycle. This paper describes the status of the modifications to the facility and briefly reviews the status of the development of the process equipment. The HFEF/S was the demonstration facility for the early Experimental Breeder Reactor II (EBR-II) melt refining/injection-casting fuel cycle. Then called the Fuel Cycle Facility, ∼400 EBR-II fuel assemblies were recycled in the two hot cells of the facility during the 1964-69 period. Since then it has been utilized as a fuels examination facility. The objective of the IFR fuel cycle program is to upgrade HFEF/S to current standards, install new process equipment, and demonstrate the commercial feasibility of the IFR pyroprocess fuel cycle

Nuclear-fuel-cycle costs. Consolidated Fuel-Reprocessing Program

International Nuclear Information System (INIS)

Burch, W.D.; Haire, M.J.; Rainey, R.H.

1981-01-01

The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel-cycle costs are given for the pressurized-water reactor once-through and fuel-recycle systems, and for the liquid-metal fast-breeder-reactor system. These calculations show that fuel-cycle costs are a small part of the total power costs. For breeder reactors, fuel-cycle costs are about half that of the present once-through system. The total power cost of the breeder-reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment

Testing and Analysis of a Composite Non-Cylindrical Aircraft Fuselage Structure . Part II; Severe Damage

Science.gov (United States)

Przekop, Adam; Jegley, Dawn C.; Lovejoy, Andrew E.; Rouse, Marshall; Wu, Hsi-Yung T.

2016-01-01

The Environmentally Responsible Aviation Project aimed to develop aircraft technologies enabling significant fuel burn and community noise reductions. Small incremental changes to the conventional metallic alloy-based 'tube and wing' configuration were not sufficient to achieve the desired metrics. One airframe concept identified by the project as having the potential to dramatically improve aircraft performance was a composite-based hybrid wing body configuration. Such a concept, however, presented inherent challenges stemming from, among other factors, the necessity to transfer wing loads through the entire center fuselage section which accommodates a pressurized cabin confined by flat or nearly flat panels. This paper discusses a finite element analysis and the testing of a large-scale hybrid wing body center section structure developed and constructed to demonstrate that the Pultruded Rod Stitched Efficient Unitized Structure concept can meet these challenging demands of the next generation airframes. Part II of the paper considers the final test to failure of the test article in the presence of an intentionally inflicted severe discrete source damage under the wing up-bending loading condition. Finite element analysis results are compared with measurements acquired during the test and demonstrate that the hybrid wing body test article was able to redistribute and support the required design loads in a severely damaged condition.

The year 2013 in the European Heart Journal--Cardiovascular Imaging: Part II.

Science.gov (United States)

Plein, Sven; Edvardsen, Thor; Pierard, Luc A; Saraste, Antti; Knuuti, Juhani; Maurer, Gerald; Lancellotti, Patrizio

2014-08-01

The new multi-modality cardiovascular imaging journal, European Heart Journal - Cardiovascular Imaging, was created in 2012. Here we summarize the most important studies from the journal's second year in two articles. Part I of the review has summarized studies in myocardial function, myocardial ischaemia, and emerging techniques in cardiovascular imaging. Part II is focussed on valvular heart diseases, heart failure, cardiomyopathies, and congenital heart diseases. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2014. For permissions please email: journals.permissions@oup.com.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

Subseabed disposal program annual report, January-December 1979. Volume II. Appendices (principal investigator progress reports). Part 1 of 2

International Nuclear Information System (INIS)

Talbert, D.M.

1981-04-01

Volume II of the sixth annual report describing the progress and evaluating the status of the Subseabed Disposal Program contains the appendices referred to in Volume I, Summary and Status. Because of the length of Volume II, it has been split into two parts for publication purposes. Part 1 contains Appendices A-O; Part 2 contains Appendices P-FF. Separate abstracts have been prepared of each Appendix for inclusion in the Energy Data Base

A state of the art on metallic fuel technology development

International Nuclear Information System (INIS)

Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh

1997-01-01

Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960's, the development of metallic fuels continued throughout the 1970's at ANL's experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980's, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab

Post-irradiation examinations on the KNK II/1 fuel element NY-203 with 400 equivalent full-power days residence time and 10 % burnup

International Nuclear Information System (INIS)

Patzer, G.; Geier, F.

1984-09-01

The fuel assembly NY-203 has been irradiated in the first core of KNK II up to a burnup of about 10 % and a residence time of 400 equivalent full-power days. The assembly contained 211 fuel pins with 6.0 mm outer diameter and fuel pellets with the composition (U 0 .7Pu 0 .3)O 2 .00. The cladding material was the austenitic steel 1.4988 lg. Some selected pins were examined in the hot cells of the KfK Karlsruhe. The post-irradiation examinations did not reveal any critical design aspects [de

An improved crude oil atmospheric distillation process for energy integration: Part II: New approach for energy saving by use of residual heat

International Nuclear Information System (INIS)

Benali, Tahar; Tondeur, Daniel; Jaubert, Jean Noël

2012-01-01

In Part I of this paper, it was shown on thermodynamic grounds that introducing a flash in the preheating train of an atmospheric oil distillation process, together with an appropriate introduction of the resulting vapour into the column, could potentially bring substantial energy savings, by reducing the duty of the preheating furnace, by doing some pre-fractionation and by reducing the column irreversibilities. Part II expands on this idea by showing how this can be done while keeping the throughput and the product characteristics unchanged. The outcome is that placing several flashes after the heat exchangers and feeding the corresponding vapour streams to the appropriate trays of the column reduces the pumparound flows and the heat brought to the preheating train. The resulting heat deficit may then be compensated in an additional heat exchanger by using low level heat recuperated from the products of the distillation and/or imported from other processes. The use of this residual heat reduces the furnace duty by approximately an equivalent amount. Thus high level energy (fuel-gas burnt in the furnace) is replaced by residual low level heat. The simulation with an example flowsheet shows that the savings on fuel could be as high as 21%. - Highlights: ► Flash installation in the preheating train of the crude oil distillation process. ► Pumparound streams and heat sent to the preheating train are reduced. ► A high level heat deficit is induced and replaced by low level heat. ► Considerable energy savings and greenhouse gas emissions are achieved.

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

Bloqueio do nervo supraescapular: procedimento importante na prática clínica. Parte II Suprascapular nerve block: important procedure in clinical practice. Part II

Directory of Open Access Journals (Sweden)

Marcos Rassi Fernandes

2012-08-01

Full Text Available O bloqueio do nervo supraescapular é um método de tratamento reprodutível, confiável e extremamente efetivo no controle da dor no ombro. Esse método tem sido amplamente utilizado por profissionais na prática clínica, como reumatologistas, ortopedistas, neurologistas e especialistas em dor, na terapêutica de enfermidades crônicas, como lesão irreparável do manguito rotador, artrite reumatoide, sequelas de AVC e capsulite adesiva, o que justifica a presente revisão (Parte II. O objetivo deste estudo foi descrever as técnicas do procedimento e suas complicações descritas na literatura, já que a primeira parte reportou as indicações clínicas, drogas e volumes utilizados em aplicação única ou múltipla. Apresentamse, detalhadamente, os acessos para a realização do procedimento tanto direto como indireto, anterior e posterior, lateral e medial, e superior e inferior. Diversas são as opções para se realizar o bloqueio do nervo supraescapular. Apesar de raras, as complicações podem ocorrer. Quando bem indicado, este método deve ser considerado.The suprascapular nerve block is a reproducible, reliable, and extremely effective treatment method in shoulder pain control. This method has been widely used by professionals in clinical practice such as rheumatologists, orthopedists, neurologists, and pain specialists in the treatment of chronic diseases such as irreparable rotator cuff injury, rheumatoid arthritis, stroke sequelae, and adhesive capsulitis, which justifies the present review (Part II. The objective of this study was to describe the techniques and complications of the procedure described in the literature, as the first part reported the clinical indications, drugs, and volumes used in single or multiple procedures. We present in details the accesses used in the procedure: direct and indirect, anterior and posterior, lateral and medial, upper and lower. There are several options to perform suprascapular nerve block

Fuel mechanical design as a boundary condition for fuel management optimization

International Nuclear Information System (INIS)

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

1988-01-01

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

Hybrid fuel cell/diesel generation total energy system, part 2

Science.gov (United States)

Blazek, C. F.

1982-11-01

Meeting the Goldstone Deep Space Communications Complex (DGSCC) electrical and thermal requirements with the existing system was compared with using fuel cells. Fuel cell technology selection was based on a 1985 time frame for installation. The most cost-effective fuel feedstock for fuel cell application was identified. Fuels considered included diesel oil, natural gas, methanol and coal. These fuel feedstocks were considered not only on the cost and efficiency of the fuel conversion process, but also on complexity and integration of the fuel processor on system operation and thermal energy availability. After a review of fuel processor technology, catalytic steam reformer technology was selected based on the ease of integration and the economics of hydrogen production. The phosphoric acid fuel cell was selected for application at the GDSCC due to its commercial readiness for near term application. Fuel cell systems were analyzed for both natural gas and methanol feedstock. The subsequent economic analysis indicated that a natural gas fueled system was the most cost effective of the cases analyzed.

Recovery of Navy distillate fuel from reclaimed product. Volume II. Literature review

Energy Technology Data Exchange (ETDEWEB)

Brinkman, D.W.; Whisman, M.L.

1984-11-01

In an effort to assist the Navy to better utilize its waste hydrocarbons, NIPER, with support from the US Department of Energy, is conducting research designed to ultimately develop a practical technique for converting Reclaimed Product (RP) into specification Naval Distillate Fuel (F-76). This first phase of the project was focused on reviewing the literature and available information from equipment manufacturers. The literature survey has been carefully culled for methodology applicable to the conversion of RP into diesel fuel suitable for Navy use. Based upon the results of this study, a second phase has been developed and outlined in which experiments will be performed to determine the most practical recycling technologies. It is realized that the final selection of one particular technology may be site-specific due to vast differences in RP volume and available facilities. A final phase, if funded, would involve full-scale testing of one of the recommended techniques at a refueling depot. The Phase I investigations are published in two volumes. Volume 1, Technical Discussion, includes the narrative and Appendices I and II. Appendix III, a detailed Literature Review, includes both a narrative portion and an annotated bibliography containing about 800 references and abstracts. This appendix, because of its volume, has been published separately as Volume 2.

EOSLT Consortium Biomass Co-firing. WP 4. Biomass co-firing in oxy-fuel combustion. Part 1. Lab- Scale Experiments

Energy Technology Data Exchange (ETDEWEB)

Fryda, L.E. [ECN Biomass, Coal and Environmental Research, Petten (Netherlands)

2011-07-15

In the frame of WP4 of the EOS LT Co-firing program, the ash formation and deposition of selected coal/biomass blends under oxyfuel and air conditions were studied experimentally in the ECN lab scale coal combustor (LCS). The fuels used were Russian coal, South African coal and Greek Lignite, either combusted separately or in blends with cocoa and olive residue. The first trial period included tests with the Russian and South African coals and their blends with cocoa, the second trial period included Lignite with olive residue tests and a final period firing only Lignite and Russian coal, mainly to check and verify the observed results. During the testing, also enriched air combustion was applied, in order to establish conclusions whether a systematic trend on ash formation and deposition exists, ranging from conventional air, to enriched air (improving post combustion applications) until oxyfuel conditions. A horizontal deposition probe equipped with thermocouples and heat transfer sensors for on line data acquisition, and a cascade impactor (staged filter) to obtain size distributed ash samples including the submicron range at the reactor exit were used. The deposition ratio and the deposition propensity measured for the various experimental conditions were higher in all oxyfuel cases. No significant variations in the ash formation mechanisms and the ash composition were established. Finally the data obtained from the tests performed under air and oxy-fuel conditions were utilised for chemical equilibrium calculations in order to facilitate the interpretation of the measured data; the results indicate that temperature dependence and fuels/blends ash composition are the major factors affecting gaseous compound and ash composition rather than the combustion environment, which seems to affect neither the ash and fine ash (submicron) formation, nor the ash composition. The ash deposition mechanisms were studied in more detail in Part II of this report.

MICROBIAL FUEL CELL

DEFF Research Database (Denmark)

2008-01-01

A novel microbial fuel cell construction for the generation of electrical energy. The microbial fuel cell comprises: (i) an anode electrode, (ii) a cathode chamber, said cathode chamber comprising an in let through which an influent enters the cathode chamber, an outlet through which an effluent...

Correlation of creep and swelling with fuel pin performance

International Nuclear Information System (INIS)

Jackson, R.J.; Washburn, D.F.; Garner, F.A.; Gilbert, E.R.

1975-09-01

The HEDL PNL-11 experiment described was one in a series of fueled subassemblies irradiated in EBR-II to demonstrate the adequacy of the FFTF fuel pin design. The cladding material, dimensions, and fuel density are prototypic of FFTF. Because neutron flux in EBR-II is lower than in FFTF, the uranium enrichment is higher in these experimental fuel pins, irradiated in EBR-II, than the FFTF enrichment for comparable linear heat rates. Some pertinent oprating conditions for the center fuel pin in this experiment are listed. This 37-pin subassembly represents, at 110,000 MWd/MTM, the highest burnup yet attained by a prototypic FFTF subassembly. Similarly, this is the highest fluence presently attained by prototypic fuel pins. A cladding breach occurred in one fuel pin which is presently being examined. Results are presented and discussed

KAFEPA-II program users' manual and description

International Nuclear Information System (INIS)

Suk, H. C.; Hwang, W.; Kim, B. G.; Sim, K. S.; Heo, Y. H.; Byun, T. S.; Park, G. S.

1992-04-01

KAFEPA-II is a computer program for simulating the behaviour of UO 2 fuel elements under normal operating conditions of a CANDU reactor. It computes the one-dimensional temperature distribution and thermal expansion of the fuel pellets. The amount of gas released during irradiation of the fuel is also computed. Thermal expansion and gas pressure inside the fuel element are then used to compute the strains and stresses in the sheath. This document is intended as a user's manual and description for KAFEPA-II. (Author)

Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

International Nuclear Information System (INIS)

Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

1982-01-01

The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

Combustion chemistry and flame structure of furan group biofuels using molecular-beam mass spectrometry and gas chromatography – Part II: 2-Methylfuran

Science.gov (United States)

Tran, Luc-Sy; Togbé, Casimir; Liu, Dong; Felsmann, Daniel; Oßwald, Patrick; Glaude, Pierre-Alexandre; Fournet, René; Sirjean, Baptiste; Battin-Leclerc, Frédérique; Kohse-Höinghaus, Katharina

2013-01-01

This is Part II of a series of three papers which jointly address the combustion chemistry of furan and its alkylated derivatives 2-methylfuran (MF) and 2,5-dimethylfuran (DMF) under premixed low-pressure flame conditions. Some of them are considered to be promising biofuels. With furan as a common basis studied in Part I of this series, the present paper addresses two laminar premixed low-pressure (20 and 40 mbar) flat argon-diluted (50%) flames of MF which were studied with electron-ionization molecular-beam mass spectrometry (EI-MBMS) and gas chromatography (GC) for equivalence ratios φ=1.0 and 1.7, identical conditions to those for the previously reported furan flames. Mole fractions of reactants, products as well as stable and reactive intermediates were measured as a function of the distance above the burner. Kinetic modeling was performed using a comprehensive reaction mechanism for all three fuels given in Part I and described in the three parts of this series. A comparison of the experimental results and the simulation shows reasonable agreement, as also seen for the furan flames in Part I before. This set of experiments is thus considered to be a valuable additional basis for the validation of the model. The main reaction pathways of MF consumption have been derived from reaction flow analyses, and differences to furan combustion chemistry under the same conditions are discussed. PMID:24518895

Nuclear fuel for light water reactors. Part 2 and conclusion

International Nuclear Information System (INIS)

1983-01-01

The article gives brief descriptions of a new cycle for nuclear fuel in the core and, in particular, fuel replacement, stock pool management for irradiated fuel elements, transport containers for irradiated nuclear fuels, treatment of low activity waste, the Climax system for long-term stocking of irradiated fuel, and transport of irradiated fuel over the Nevada Test Site. (A.E.W.)

KUCA critical experiments using MEU fuel (II)

Energy Technology Data Exchange (ETDEWEB)

Kanda, Keiji; Hayashi, Masatoshi; Shiroya, Seiji; Kobayashi, Keiji; Fukui, Hiroshi; Mishima, Kaichiro; Shibata, Toshikazu [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka (Japan)

1983-09-01

Due to mutual concerns in the USA and Japan about the proliferation potential of highly-enriched uranium (HEU), a joint study program I was initiated between Argonne National Laboratory (ANL and Kyoto University Research Reactor Institute (KURRI) in 1978. In accordance with the reduced enrichment for research and test reactor (RERTR) program, the alternatives were studied for reducing the enrichment of the fuel to be used in the Kyoto University High Flux Reactor (KUHFR). The KUHFR has a distinct feature in its core configuration it is a coupled-core. Each annular shaped core is light-water-moderated and placed within a heavy water reflector with a certain distance between them. The phase A reports of the joint ANL-KURRI program independently prepared by two laboratories in February 1979, 3,4 concluded that the use of medium-enrichment uranium (MEU, 45%) in the KUHFR is feasible, pending results of the critical experiments in the Kyoto University Critical Assembly (KUCA) 5 and of the burnup test in the Oak Ridge Research Reactor 6 (ORR). An application of safety review (Reactor Installation License) for MEU fuel to be used in the KUCA was submitted to the Japanese Government in March 1980, and a license was issued in August 1980. Subsequently, the application for 'Authorization before Construction' was submitted and was authorized in September 1980. Fabrication of MEU fuel-elements for the KUCA experiments by CERCA in France was started in September 1980, and was completed in March 1981. The critical experiments in the KUCA with MEU fuel were started on a single-core in May 1981 as a first step. The first critical state of the core using MEU fuel was achieved at 312 p.m. in May 12, 1981. After that, the reactivity effects of the outer side-plates containing boron burnable poison were measured. At Munich Meeting in Sept., 1981, we presented a paper on critical mass and reactivity of burnable poison in the MEU core. Since then we carried out the following experiments

KUCA critical experiments using MEU fuel (II)

International Nuclear Information System (INIS)

Kanda, Keiji; Hayashi, Masatoshi; Shiroya, Seiji; Kobayashi, Keiji; Fukui, Hiroshi; Mishima, Kaichiro; Shibata, Toshikazu

1983-01-01

Due to mutual concerns in the USA and Japan about the proliferation potential of highly-enriched uranium (HEU), a joint study program I was initiated between Argonne National Laboratory (ANL and Kyoto University Research Reactor Institute (KURRI) in 1978. In accordance with the reduced enrichment for research and test reactor (RERTR) program, the alternatives were studied for reducing the enrichment of the fuel to be used in the Kyoto University High Flux Reactor (KUHFR). The KUHFR has a distinct feature in its core configuration it is a coupled-core. Each annular shaped core is light-water-moderated and placed within a heavy water reflector with a certain distance between them. The phase A reports of the joint ANL-KURRI program independently prepared by two laboratories in February 1979, 3,4 concluded that the use of medium-enrichment uranium (MEU, 45%) in the KUHFR is feasible, pending results of the critical experiments in the Kyoto University Critical Assembly (KUCA) 5 and of the burnup test in the Oak Ridge Research Reactor 6 (ORR). An application of safety review (Reactor Installation License) for MEU fuel to be used in the KUCA was submitted to the Japanese Government in March 1980, and a license was issued in August 1980. Subsequently, the application for 'Authorization before Construction' was submitted and was authorized in September 1980. Fabrication of MEU fuel-elements for the KUCA experiments by CERCA in France was started in September 1980, and was completed in March 1981. The critical experiments in the KUCA with MEU fuel were started on a single-core in May 1981 as a first step. The first critical state of the core using MEU fuel was achieved at 312 p.m. in May 12, 1981. After that, the reactivity effects of the outer side-plates containing boron burnable poison were measured. At Munich Meeting in Sept., 1981, we presented a paper on critical mass and reactivity of burnable poison in the MEU core. Since then we carried out the following experiments

Modeling of constituent redistribution in U-Pu-Zr metallic fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)]. E-mail: yskim@anl.gov; Hayes, S.L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Yacout, A.M. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2006-12-01

A computer model was developed to analyze constituent redistribution in U-Pu-Zr metallic nuclear fuels. Diffusion and thermochemical properties were parametrically determined to fit the postirradiation data from a fuel test performed in the Experimental Breeder Reactor II (EBR-II). The computer model was used to estimate redistribution profiles of fuels proposed for the conceptual designs of small modular fast reactors. The model results showed that the level of redistribution of the fuel constituents of the designs was similar to the measured data from EBR-II.

Reactivity feedback components of a homogeneous U10Zr-fueled 900 MWt LMR

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1988-01-01

The linear and Doppler feedback components of the regional contributions of the power-reactivity-decrement (PRD) and of the temperature coefficient of reactivity for a 900 MWt homogeneous U10Zr-fueled sodium-cooled reactor are calculated. The PRD components are separated into power dependent and power-to-flow dependent parts. The values of PRD and temperature coefficient components are compared with corresponding quantities calculated for the Experimental Breeder Reactor II. The implications of these comparisons upon inherent safety characteristics of metal-fueled sodium-cooled reactors are discussed

Hot Fuel Examination Facility/South

Energy Technology Data Exchange (ETDEWEB)

1990-05-01

This document describes the potential environmental impacts associated with proposed modifications to the Hot Fuel Examination Facility/South (HFEF/S). The proposed action, to modify the existing HFEF/S at the Argonne National Laboratory-West (ANL-W) on the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, would allow important aspects of the Integral Fast Reactor (IFR) concept, offering potential advantages in nuclear safety and economics, to be demonstrated. It would support fuel cycle experiments and would supply fresh fuel to the Experimental Breeder Reactor-II (EBR-II) at the INEL. 35 refs., 12 figs., 13 tabs.

Hot Fuel Examination Facility/South

International Nuclear Information System (INIS)

1990-05-01

This document describes the potential environmental impacts associated with proposed modifications to the Hot Fuel Examination Facility/South (HFEF/S). The proposed action, to modify the existing HFEF/S at the Argonne National Laboratory-West (ANL-W) on the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, would allow important aspects of the Integral Fast Reactor (IFR) concept, offering potential advantages in nuclear safety and economics, to be demonstrated. It would support fuel cycle experiments and would supply fresh fuel to the Experimental Breeder Reactor-II (EBR-II) at the INEL. 35 refs., 12 figs., 13 tabs

Three Mile Island: a report to the commissioners and to the public. Volume II, Part 3

International Nuclear Information System (INIS)

1979-01-01

This is the third and final part of the second volume of a study of the Three Mile Island accident. Part 3 of Volume II contains descriptions and assessments of responses to the accident by the utility and by the NRC and other government agencies

Pelletised fuel production from coal tailings and spent mushroom compost - Part II. Economic feasibility based on cost analysis

International Nuclear Information System (INIS)

Ryu, Changkook; Khor, Adela; Sharifi, Vida N.; Swithenbank, Jim

2008-01-01

Due to the growing market for sustainable energy, in order to increase the quality of the fuels, pellets are being produced from various materials such as wood and other biomass energy crops, and municipal waste. This paper presents the results from an economic feasibility study for pellet production using blends of two residue materials: coal tailings from coal cleaning and spent mushroom compost (SMC) from mushroom production. Key variables such as the mixture composition, raw material haulage and plant scale were considered and the production costs were compared to coal and biomass energy prices. For both wet materials, the moisture content was the critical parameter that influenced the fuel energy costs. The haulage distance of the raw materials was another factor that can pose a high risk. The results showed that the pellet production from the above two materials can be viable when a less energy-intensive drying process is utilised. Potential market outlets and ways to lower the costs are also discussed in this paper. (author)

A state of the art on metallic fuel technology development

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960`s, the development of metallic fuels continued throughout the 1970`s at ANL`s experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980`s, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab.

Estimation of energetic efficiency of heat supply in front of the aircraft at supersonic accelerated flight. Part II. Mathematical model of the trajectory boost part and computational results

Science.gov (United States)

Latypov, A. F.

2009-03-01

The fuel economy was estimated at boost trajectory of aerospace plane during energy supply to the free stream. Initial and final velocities of the flight were given. A model of planning flight above cold air in infinite isobaric thermal wake was used. The comparison of fuel consumption was done at optimal trajectories. The calculations were done using a combined power plant consisting of ramjet and liquid-propellant engine. An exergy model was constructed in the first part of the paper for estimating the ramjet thrust and specific impulse. To estimate the aerodynamic drag of aircraft a quadratic dependence on aerodynamic lift is used. The energy for flow heating is obtained at the sacrifice of an equivalent decrease of exergy of combustion products. The dependencies are obtained for increasing the range coefficient of cruise flight at different Mach numbers. In the second part of the paper, a mathematical model is presented for the boost part of the flight trajectory of the flying vehicle and computational results for reducing the fuel expenses at the boost trajectory at a given value of the energy supplied in front of the aircraft.

DETAILS OF OPERATIONS PERFORMED BY THE REMOTE CONTROL ROBOT (CONCEPT TO THE HORIZONTAL FUEL CHANNEL DURING DECOMMISSIONING PHASE OF NUCLEAR REACTOR CALANDRIA STRUCTURE. PART II: INSIDE OPERATIONS

Directory of Open Access Journals (Sweden)

Constantin POPESCU

2017-05-01

Full Text Available The authors contribution to this paper is to present a concept solution of a remote control robot (RCR used for decommissioning of the horizontal fuel channels pressure tube in the CANDU nuclear reactor. In this paper the authors highlight few details of geometry, operations, constraints by kinematics and dynamics of the robot movement inside of the reactor fuel channel. Inside operations performed has as the main steps of dismantling process the followings: unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, cut the ends of the pressure tube, extract the pressure tube and cut it in small parts, sorting and storage extracted items in the safe robot container. All steps are performed in automatic mode. The remote control robot (RCR represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the inside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection.

Economic Analysis of Symbiotic Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

International Nuclear Information System (INIS)

Williams, Kent Alan; Shropshire, David E.

2009-01-01

A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

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

Science.gov (United States)

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

2015-02-01

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

Numerical Simulation of Projectile Impact on Mild Steel Armour Platesusing LS-DYNA, Part II: Parametric Studies

OpenAIRE

M. Raguraman; A. Deb; N. K. Gupta; D. K. Kharat

2008-01-01

In Part I of the current two-part series, a comprehensive simulation-based study of impact of jacketed projectiles on mild steel armour plates has been presented. Using the modelling procedures developed in Part I, a number of parametric studies have been carried out for the same mild steel plates considered in Part I and reported here in Part II. The current investigation includes determination of ballistic limits of a given target plate for different projectile diameters and impact velociti...

PIC Simulations in Low Energy Part of PIP-II Proton Linac

Energy Technology Data Exchange (ETDEWEB)

Romanov, Gennady

2014-07-01

The front end of PIP-II linac is composed of a 30 keV ion source, low energy beam transport line (LEBT), 2.1 MeV radio frequency quadrupole (RFQ), and medium energy beam transport line (MEBT). This configuration is currently being assembled at Fermilab to support a complete systems test. The front end represents the primary technical risk with PIP-II, and so this step will validate the concept and demonstrate that the hardware can meet the specified requirements. SC accelerating cavities right after MEBT require high quality and well defined beam after RFQ to avoid excessive particle losses. In this paper we will present recent progress of beam dynamic study, using CST PIC simulation code, to investigate partial neutralization effect in LEBT, halo and tail formation in RFQ, total emittance growth and beam losses along low energy part of the linac.

Estimating fuel cycle externalities: Analytical methods and issues. Report number 2 on the external costs and benefits of fuel cycles: A study by the U.S. Department of Energy and the Commission of the European Communities

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-07-01

This report, the second in a series of eight reports, is part of a joint study by the U.S. Department of Energy (DOE) and the Commission of the European Communities (EC) 'on the externalities of fuel cycles.' Part I illustrates the use of the atmospheric dispersion and transformation modeling that this study recommends for airborne pollutants in the coal, biomass, oil, and natural gas fuel cycles. Part II of this volume contains a paper which reviews the scientific literature on ecological impacts associated with power plant discharges. Part III contains papers summarizing the relevant health effects literature. Part IV contains papers on methods of economic evaluation. Part V contains four papers on various issues related to the estimation of externalities and their use in public policy. The final part is Part VI, and it contains a paper which describes a system for summarizing analysts' assessments of the quality of the information that an analysis uses to estimate externalities. This system allows analysts to provide information, not only on their best estimates, but also on a range of estimates, on uncertainty, on the quality of the data, and on other factors that better reflect the full dimension of making estimates under uncertainty. The system has broad applicability beyond fuel cycle externalities, as well.

Estimating fuel cycle externalities: Analytical methods and issues. Report number 2 on the external costs and benefits of fuel cycles: A study by the U.S. Department of Energy and the Commission of the European Communities

International Nuclear Information System (INIS)

1994-07-01

This report, the second in a series of eight reports, is part of a joint study by the U.S. Department of Energy (DOE) and the Commission of the European Communities (EC) 'on the externalities of fuel cycles.' Part I illustrates the use of the atmospheric dispersion and transformation modeling that this study recommends for airborne pollutants in the coal, biomass, oil, and natural gas fuel cycles. Part II of this volume contains a paper which reviews the scientific literature on ecological impacts associated with power plant discharges. Part III contains papers summarizing the relevant health effects literature. Part IV contains papers on methods of economic evaluation. Part V contains four papers on various issues related to the estimation of externalities and their use in public policy. The final part is Part VI, and it contains a paper which describes a system for summarizing analysts' assessments of the quality of the information that an analysis uses to estimate externalities. This system allows analysts to provide information, not only on their best estimates, but also on a range of estimates, on uncertainty, on the quality of the data, and on other factors that better reflect the full dimension of making estimates under uncertainty. The system has broad applicability beyond fuel cycle externalities, as well

EBR-II blanket fuel leaching test using simulated J-13 well water.

Energy Technology Data Exchange (ETDEWEB)

Fonnesbeck, J. E.

1998-05-15

A pulsed-flow leaching test is being conducted using three EBR-II blanket fuel segments. These samples are immersed in simulated J-13 well water. The samples are kept at a constant temperature of 90 C. Leachate is exchanged weekly and analyzed for various nuclides which are of interest from a mobility and longevity point of view. Our primary interest is in the longer-lived species such as {sup 99}Tc, {sup 237}Np, and {sup 241}Am. In addition, the behavior of U, Pu, {sup 90}Sr, and {sup 137}Cs are being analyzed. During the course of this experiment, an interesting observation has been made involving one of the samples which could indicate the possible rapid ''anoxic'' oxidation of uranium metal to UO{sub 2}.

Fuel Modelling at Extended Burnup (Fumex-II). Report of a Coordinated Research Project 2002-2007

International Nuclear Information System (INIS)

2012-08-01

to fuel licensing. This report describes the results of the coordinated research project on fuel modelling at extended burnup (FUMEX-II). This programme was initiated in 2000 and completed in 2006. It followed previous programmes on fuel modelling, D-COM which was conducted between 1982 and 1984, and the FUMEX programme which was conducted between 1993 and 1996. The participants used a mixture of data, derived from actual irradiation histories, in particular those with PIE measurements from high burnup commercial and experimental fuels, combined with idealized power histories intended to represent possible future extended dwell, commercial irradiations, to test code capabilities at high burnup. All participants have carried out calculations on the six priority cases selected from the 27 cases identified to them at the first research coordination meeting (RCM). At the second RCM, three further priority cases were identified and have been modelled. These priority cases have been chosen as the best available to help determine which of the many high burnup models used in the codes best reflect reality. The participants are using the remaining cases for verification and validation purposes as well as inter-code comparisons. The codes participating in the exercise have been developed for a wide variety of purposes, including predictions for fuel operation in PWR, BWR, WWER, the pressurized HWR type, CANDU and other reactor types. They are used as development tools as well as for routine licensing calculations, where code configuration is strictly controlled.

Engineering study: Fast Flux Test Facility fuel reprocessing

International Nuclear Information System (INIS)

Beary, M.M.; Raab, G.J.; Reynolds, W.R. Jr.; Yoder, R.A.

1974-01-01

Several alternatives were studied for reprocessing FFTF fuels at Hanford. Alternative I would be to decontaminate and trim the fuel at T Plant and electrolytically dissolve the fuel at Purex. Alternative II would be to decontaminate and shear leach the fuels in a new facility near Purex. Alternative III would be to decontaminate and store fuel elements indefinitely at T Plant for subsequent offsite shipment. Alternative I, 8 to 10 M$ and 13 quarter-years; for Alternative II, 24 to 28 M$ and 20 quarter-years; for Alternative III, 3 to 4 M$ and 8 quarter-years. Unless there is considerable slippage in the FFTF shipping schedule, it would not be possible to build a new facility as described in Alternative II in time without building temporary storage facilities at T Plant, as described in Alternative III

The EBR-II X501 Minor Actinide Burning Experiment

Energy Technology Data Exchange (ETDEWEB)

W. J. Carmack; M. K. Meyer; S. L. Hayes; H. Tsai

2008-01-01

The X501 experiment was conducted in EBR II as part of the Integral Fast Reactor program to demonstrate minor actinide burning through the use of a homogeneous recycle scheme. The X501 subassembly contained two metallic fuel elements loaded with relatively small quantities of americium and neptunium. Interest in the behavior of minor actinides (MA) during fuel irradiation has prompted further examination of existing X501 data and generation of new data where needed in support of the U.S. waste transmutation effort. The X501 experiment is one of the few MA bearing fuel irradiation tests conducted worldwide, and knowledge can be gained by understanding the changes in fuel behavior due to addition of MAs. Of primary interest are the effect of the MAs on fuel cladding chemical interaction and the redistribution behavior of americium. The quantity of helium gas release from the fuel and any effects of helium on fuel performance are also of interest. It must be stressed that information presented at this time is based on the limited PIE conducted in 1995–1996 and, currently, represents a set of observations rather than a complete understanding of fuel behavior. This report provides a summary of the X501 fabrication, characterization, irradiation, and post irradiation examination.

Final project report: TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

International Nuclear Information System (INIS)

Montoya, G.M.

1993-02-01

This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project including, Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 M 3 of mixed waste

Nuclear fuels policy. Report of the Atlantic Council's Nuclear Fuels Policy Working Group

International Nuclear Information System (INIS)

Anon.

1976-01-01

This Policy Paper recommends the actions deemed necessary to assure that future U.S. and non-Communist countries' nuclear fuels supply will be adequate, considering the following: estimates of modest growth in overall energy demand, electrical energy demand, and nuclear electrical energy demand in the U.S. and abroad, predicated upon the continuing trends involving conservation of energy, increased use of electricity, and moderate economic growth (Chap. I); possibilities for the development and use of all domestic resources providing energy alternatives to imported oil and gas, consonant with current environmental, health, and safety concerns (Chap. II); assessment of the traditional energy sources which provide current alternatives to nuclear energy (Chap. II); evaluation of realistic expectations for additional future energy supplies from prospective technologies: enhanced recovery from traditional sources and development and use of oil shales and synthetic fuels from coal, fusion and solar energy (Chap. II); an accounting of established nuclear technology in use today, in particular the light water reactor, used for generating electricity (Chap. III); an estimate of future nuclear technology, in particular the prospective fast breeder (Chap. IV); current and projected nuclear fuel demand and supply in the U.S. and abroad (Chaps. V and VI); the constraints encountered today in meeting nuclear fuels demand (Chap. VII); and the major unresolved issues and options in nuclear fuels supply and use (Chap. VIII). The principal conclusions and recommendations (Chap. IX) are that the U.S. and other industrialized countries should strive for increased flexibility of primary energy fuel sources, and that a balanced energy strategy therefore depends on the secure supply of energy resources and the ability to substitute one form of fuel for another

Development of a methanol reformer for fuel cell vehicles

Energy Technology Data Exchange (ETDEWEB)

Lindstroem, Baard

2003-03-01

Vehicles powered by fuel cells are from an environmental aspect superior to the traditional automobile using internal combustion of gasoline. Power systems which are based upon fuel cell technology require hydrogen for operation. The ideal fuel cell vehicle would operate on pure hydrogen stored on-board. However, storing hydrogen on-board the vehicle is currently not feasible for technical reasons. The hydrogen can be generated on-board using a liquid hydrogen carrier such as methanol and gasoline. The objective of the work presented in this thesis was to develop a catalytic hydrogen generator for automotive applications using methanol as the hydrogen carrier. The first part of this work gives an introduction to the field of methanol reforming and the properties of a fuel cell based power system. Paper I reviews the catalytic materials and processes available for producing hydrogen from methanol. The second part of this thesis consists of an experimental investigation of the influence of the catalyst composition, materials and process parameters on the activity and selectivity for the production of hydrogen from methanol. In Papers II-IV the influence of the support, carrier and operational parameters is studied. In Paper V an investigation of the catalytic properties is performed in an attempt to correlate material properties with performance of different catalysts. In the third part of the thesis an investigation is performed to elucidate whether it is possible to utilize oxidation of liquid methanol as a heat source for an automotive reformer. In the study which is presented in Paper VI a large series of catalytic materials are tested and we were able to minimize the noble metal content making the system more cost efficient. In the final part of this thesis the reformer prototype developed in the project is evaluated. The reformer which was constructed for serving a 5 k W{sub e} fuel cell had a high performance with near 100 % methanol conversion and CO

EBR-II blanket fuel leaching test using simulated J-13 well water

International Nuclear Information System (INIS)

Fonnesbeck, J. E.

1999-01-01

This paper discusses the results of a pulsed-flow leaching test using simulated J-13 well water leachant. This test was performed on three blanket fuel segments from the ANL-W EBR-II nuclear reactor which were originally made up of depleted uranium (DU). This experiment was designed to mimic conditions which would exist if, upon disposal of this material in a geological repository, it came in direct contact with groundwater. These segments were contained in pressure vessels and maintained at a constant temperature of 90 C. Weekly aliquots of leachate were taken from the three vessels and replaced with an equal volume of fresh leachant. These weekly aliquots were analyzed for both 90 Sr and 137 Cs. The results of the pulsed-flow leach test showed the formation of uranium oxide (UO 2 ) and uranium hydride (UH 3 ) particulate with rapid release of the 137 Cs and 90 Sr to the leachant. On the fifth week of sampling, one of the vessels became over pressurized and vented gas when opened. The most reasonable explanation for the presence of gas in this vessel is that the unoxidized uranium metal in the blanket segment could have reacted with the surrounding water leachant to form hydrogen. However, an investigation is currently being undertaken to both qualify and quantify H 2 formation during uranium spent nuclear fuel corrosion in water

A thermoelectric power generating heat exchanger: Part II – Numerical modeling and optimization

DEFF Research Database (Denmark)

Sarhadi, Ali; Bjørk, Rasmus; Lindeburg, N.

2016-01-01

In Part I of this study, the performance of an experimental integrated thermoelectric generator (TEG)-heat exchanger was presented. In the current study, Part II, the obtained experimental results are compared with those predicted by a finite element (FE) model. In the simulation of the integrated...... TEG-heat exchanger, the thermal contact resistance between the TEG and the heat exchanger is modeled assuming either an ideal thermal contact or using a combined Cooper–Mikic–Yovanovich (CMY) and parallel plate gap formulation, which takes into account the contact pressure, roughness and hardness...

Evolution of fuel rod support under irradiation consequences on the mechanical behavior of fuel assembly

International Nuclear Information System (INIS)

Billerey, A.; Bouffioux, P.

2002-01-01

The complete paper follows. According to the fuel management policy in French PWR with respect to high burn-up, the prediction of the mechanical behavior of the irradiated fuel assembly is required as far as excessive deformations of fuel assembly might lead to incomplete Rod Cluster Control Assembly insertion (safety problems) and fretting wear lead to leaking rods (plant operation problems). One of the most important parameter is the evolution of the fuel rod support in the grid cell as it directly governs the mechanical behavior of the fuel assembly and consequently allows to predict the behavior of irradiated structure in terms of (i) axial and lateral deformation (global behavior of the assembly) and (ii) fretting wear (local behavior of the rod). Fuel rod support is provided by a spring-dimple system fixed on the grid. During irradiation, the spring force decreases and a gap between the rod and the spring might open. This phenomenon is due to (i) irradiation-induced stress relaxation for the spring and for the dimples, (ii) grid growth and (iii) reduction of rod diameter. Two models have been developed to predict the behavior of the rod in the grid cell. The first model is able to evaluate the spring force relaxation during irradiation. The second one is able to evaluate the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (i) the creep laws of the grid materials, (ii) the growth law of the grid, (iii) the evolution of rod diameter and (iv) the design of the fuel rod support. The objectives of this paper are to: (i) evaluate the consequences of grid support design modifications on the fretting sensitivity in terms of predicted maximum gap during irradiation and operational time to gap appearance; (ii) evaluate, using a non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the mechanical behavior of the full assembly in terms of axial and

Uncertainty and sensitivity analysis in reactivity-initiated accident fuel modeling: synthesis of organisation for economic co-operation and development (OECD/nuclear energy agency (NEA benchmark on reactivity-initiated accident codes phase-II

Directory of Open Access Journals (Sweden)

Olivier Marchand

2018-03-01

Full Text Available In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010–2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II was thus launched early in 2014, which has been organized in two complementary activities: (1 comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2 assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made. Keywords: RIA, Codes Benchmarking, Fuel Modelling, OECD

Understanding Medicines: Conceptual Analysis of Nurses' Needs for Knowledge and Understanding of Pharmacology (Part I). Understanding Medicines: Extending Pharmacology Education for Dependent and Independent Prescribing (Part II).

Science.gov (United States)

Leathard, Helen L.

2001-01-01

Part I reviews what nurses need to know about the administration and prescription of medicines. Part II addresses drug classifications, actions and effects, and interactions. Also discussed are the challenges pharmacological issues pose for nursing education. (SK)

Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

International Nuclear Information System (INIS)

Weber, J.W.; Dutt, D.S.

1976-01-01

An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

Uncertainty estimation with a small number of measurements, part II: a redefinition of uncertainty and an estimator method

Science.gov (United States)

Huang, Hening

2018-01-01

This paper is the second (Part II) in a series of two papers (Part I and Part II). Part I has quantitatively discussed the fundamental limitations of the t-interval method for uncertainty estimation with a small number of measurements. This paper (Part II) reveals that the t-interval is an ‘exact’ answer to a wrong question; it is actually misused in uncertainty estimation. This paper proposes a redefinition of uncertainty, based on the classical theory of errors and the theory of point estimation, and a modification of the conventional approach to estimating measurement uncertainty. It also presents an asymptotic procedure for estimating the z-interval. The proposed modification is to replace the t-based uncertainty with an uncertainty estimator (mean- or median-unbiased). The uncertainty estimator method is an approximate answer to the right question to uncertainty estimation. The modified approach provides realistic estimates of uncertainty, regardless of whether the population standard deviation is known or unknown, or if the sample size is small or large. As an application example of the modified approach, this paper presents a resolution to the Du-Yang paradox (i.e. Paradox 2), one of the three paradoxes caused by the misuse of the t-interval in uncertainty estimation.

Development of methods for theoretical analysis of nuclear reactors (Phase II), I-V, Part IV, Fuel depletion

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1962-10-01

This report includes the analysis of plutonium isotopes from U 238 depletion chain. Two theoretical approaches for solving the depletion of fuel are shown. One results in the system of differential equations that can be solved only by using electronic calculators and the second, Machinari-Goto method enables obtaining analytical equations for approximative values of particular nuclei. In addition, differential equations are given for different approximation levels in calculating Pu 239 , as well as relations between the released energy and irradiation [sr

AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XXIV, I--MAINTAINING THE FUEL SYSTEM PART III--CATERPILLAR DIESEL ENGINE, II--UNDERSTANDING THE VOLTAGE REGULATOR/ALTERNATOR.

Science.gov (United States)

Minnesota State Dept. of Education, St. Paul. Div. of Vocational and Technical Education.

THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL AND BATTERY CHARGING SYSTEM. TOPICS ARE (1) INJECTION TIMING CONTROLS, (2) GOVERNOR, (3) FUEL SYSTEM MAINTENANCE TIPS, (4) THE CHARGING SYSTEM, (5) REGULATING THE GENERATOR/ALTERNATOR, AND (6) CHARGING SYSTEM SERVICE…

Simulation of TRIGA Mark II Benchmark Experiment using WIMSD4 and CITATION codes

International Nuclear Information System (INIS)

Dalle, Hugo Moura; Pereira, Claubia

2000-01-01

This paper presents a simulation of the TRIGA Mark II Benchmark Experiment, Part I: Steady-State Operation and is part of the calculation methodology validation developed to the neutronic calculation of the CDTN's TRIGA IPR - R1 reactor. A version of the WIMSD4, obtained in the Centro de Tecnologia Nuclear, in Cuba, was used in the cells calculation. In the core calculations was adopted the diffusion code CITATION. Was adopted a 3D representation of the core and the calculations were carried out at two energy groups. Many of the experiments were simulated, including, K eff , control rods reactivity worth, fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or on an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental. results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or in an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution. (author)

Bayesian inference for psychology. Part II: Example applications with JASP.

Science.gov (United States)

Wagenmakers, Eric-Jan; Love, Jonathon; Marsman, Maarten; Jamil, Tahira; Ly, Alexander; Verhagen, Josine; Selker, Ravi; Gronau, Quentin F; Dropmann, Damian; Boutin, Bruno; Meerhoff, Frans; Knight, Patrick; Raj, Akash; van Kesteren, Erik-Jan; van Doorn, Johnny; Šmíra, Martin; Epskamp, Sacha; Etz, Alexander; Matzke, Dora; de Jong, Tim; van den Bergh, Don; Sarafoglou, Alexandra; Steingroever, Helen; Derks, Koen; Rouder, Jeffrey N; Morey, Richard D

2018-02-01

Bayesian hypothesis testing presents an attractive alternative to p value hypothesis testing. Part I of this series outlined several advantages of Bayesian hypothesis testing, including the ability to quantify evidence and the ability to monitor and update this evidence as data come in, without the need to know the intention with which the data were collected. Despite these and other practical advantages, Bayesian hypothesis tests are still reported relatively rarely. An important impediment to the widespread adoption of Bayesian tests is arguably the lack of user-friendly software for the run-of-the-mill statistical problems that confront psychologists for the analysis of almost every experiment: the t-test, ANOVA, correlation, regression, and contingency tables. In Part II of this series we introduce JASP ( http://www.jasp-stats.org ), an open-source, cross-platform, user-friendly graphical software package that allows users to carry out Bayesian hypothesis tests for standard statistical problems. JASP is based in part on the Bayesian analyses implemented in Morey and Rouder's BayesFactor package for R. Armed with JASP, the practical advantages of Bayesian hypothesis testing are only a mouse click away.

Fuel performance experience at TVO nuclear power plant

International Nuclear Information System (INIS)

Patrakka, E.T.

1985-01-01

TVO nuclear power plant consists of two BWR units of ASEA-ATOM design. The fuel performance experience extending through six cycles at TVO I and four cycles at TVO II is reported. The experience obtained so far is mainly based on ASEA-ATOM 8 x 8 fuel and has been satisfactory. Until autumn 1984 one leaking fuel assembly had been identified at TVO I and none at TVO II. Most of the problems encountered have been related to leaf spring screws and channel screws. The experience indicates that satisfactory fuel performance can be achieved when utilizing strict operational rules and proper control of fuel design and manufacture. (author)

Low enrichment fuel development at INEL

International Nuclear Information System (INIS)

Newton, D.G.

1993-01-01

EG and G Idaho, Inc. is under contract to the Department of Energy to operate the Idaho National Engineering Laboratory (INEL). The INEL is located in southeastern Idaho. This facility has been operating since 1949 and was originally called the National Reactor Testing Station. Several contractors manage projects on this facility. Most projects at INEL are concerned with either reactor safety or irradiation testing. At Test Area North, for example, experiments are being conducted on the effects of loss of coolant. At the Test Reactor Area the ATR (Advanced Test Reactor) and ETR (Engineering Test Reactor) are used for irradiation testing and, of course, those of you working at Argonne will recognize the Experimental Breeder Reactors I and II. SPERT is an acronym for Special Power Excursion Reactor Test. A part of this former reactor facility has been converted into a fuel fabrication laboratory facility. At SPERT IV a miniature fabrication facility has been set up to duplicate the aluminide plate fuel processing line at Atomics International. In other words, a model of the supplier's processing has been created, so that what process changes are developed here can then be scaled up to production. The process is described showing: making UAI x powder, making compact for fuel core, making experimental fuel plate and compact assembly, inspection and testing the fuel plate. Main concern was related to possible swelling

Direct electrical heating of irradiated metal fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

1985-01-01

The Integral Fast Reactor (IFR) concept proposed by Argonne National Laboratory utilizes a metal fuel core. Reactor safety analysis requires information on the potential for fuel axial expansion during severe thermal transients. In addition to a comparatively large thermal expansion coefficient, metallic fuel has a unique potential for enhanced pre-failure expansion driven by retained fission gas and ingested bond sodium. In this paper, the authors present preliminary results from three direct electrical heating (DEH) experiments performed on irradiated metal fuel to investigate axial expansion behavior. The test samples were from Experimental Breeder Reactor II (EBR-II) driver fuel ML-11 irradiated to 8 at.% burnup. Preliminary analysis of the results suggest that enhanced expansion driven by trapped fission gas can occur

Scope Oriented Thermoeconomic analysis of energy systems. Part II: Formation Structure of Optimality for robust design

International Nuclear Information System (INIS)

Piacentino, Antonio; Cardona, Ennio

2010-01-01

This paper represents the Part II of a paper in two parts. In Part I the fundamentals of Scope Oriented Thermoeconomics have been introduced, showing a scarce potential for the cost accounting of existing plants; in this Part II the same concepts are applied to the optimization of a small set of design variables for a vapour compression chiller. The method overcomes the limit of most conventional optimization techniques, which are usually based on hermetic algorithms not enabling the energy analyst to recognize all the margins for improvement. The Scope Oriented Thermoeconomic optimization allows us to disassemble the optimization process, thus recognizing the Formation Structure of Optimality, i.e. the specific influence of any thermodynamic and economic parameter in the path toward the optimal design. Finally, the potential applications of such an in-depth understanding of the inner driving forces of the optimization are discussed in the paper, with a particular focus on the sensitivity analysis to the variation of energy and capital costs and on the actual operation-oriented design.

An alternative LEU design for the FRM-II

International Nuclear Information System (INIS)

Hanan, N.A.; Mo, S.C.; Smith, R.S.; Matos, J.E.

1997-02-01

The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm[sup 3] and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance (8 x 10[sup 14] n/cm[sup 2]/s in the reflector). LEU silicide fuel with 4.5 g/cm[sup 3] has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Computer models for the HEU and LEU designs have been exchanged between TUM and ANL and discrepancies have been resolved. The following issues are addressed: qualification of HEU and LEU silicide fuels, stability of the fuel plates, gamma heating in the heavy water reflector, a hypothetical accident involving the configuration of the reflector, a loss of primary coolant flow transient due to an interrupted power supply, the radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. Calculations were also done to address the possibility that new high density LEU fuels could be developed that would allow conversion of the TUM HEU design to LEU fuel. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility

Structure Learning and Statistical Estimation in Distribution Networks - Part II

Energy Technology Data Exchange (ETDEWEB)

Deka, Deepjyoti [Univ. of Texas, Austin, TX (United States); Backhaus, Scott N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chertkov, Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-02-13

Limited placement of real-time monitoring devices in the distribution grid, recent trends notwithstanding, has prevented the easy implementation of demand-response and other smart grid applications. Part I of this paper discusses the problem of learning the operational structure of the grid from nodal voltage measurements. In this work (Part II), the learning of the operational radial structure is coupled with the problem of estimating nodal consumption statistics and inferring the line parameters in the grid. Based on a Linear-Coupled(LC) approximation of AC power flows equations, polynomial time algorithms are designed to identify the structure and estimate nodal load characteristics and/or line parameters in the grid using the available nodal voltage measurements. Then the structure learning algorithm is extended to cases with missing data, where available observations are limited to a fraction of the grid nodes. The efficacy of the presented algorithms are demonstrated through simulations on several distribution test cases.

Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Choi, Hang Bok

2005-03-01

Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

Emerging Forms of the Part II of Jonathan Swift's Novel “Gulliver’s Travels”

Directory of Open Access Journals (Sweden)

Svitlana Tikhonenko

2017-11-01

Full Text Available The article is devoted to the study of grotesque forms in Jonathan Swift's novel "Gulliver’s Travels" based on the text of part II of the novel "A Voyage to Brobdingnag". On the basis of the selected actual material, displays of the grotesque elements in the semantic field of the work’s text are traced. The grotesque world of the novel is the author's model of mankind, in which J. Swift presents his view not only on the state of the modern system of England, but also on the nature of man in general, reveals the peculiarities of the psychology of human nature, especially human socialization. In part II, the author continues to develop a complex and contradictory picture of human existence in front of the reader, the world of giants appears as an ambivalent system in which the features of an ideal society and ideal ruler, in author’s opinion, with the ugly face of man and society, are marvelously combined.

Carbon deposition thresholds on nickel-based solid oxide fuel cell anodes II. Steam:carbon ratio and current density

Science.gov (United States)

Kuhn, J.; Kesler, O.

2015-03-01

For the second part of a two part publication, coking thresholds with respect to molar steam:carbon ratio (SC) and current density in nickel-based solid oxide fuel cells were determined. Anode-supported button cell samples were exposed to 2-component and 5-component gas mixtures with 1 ≤ SC ≤ 2 and zero fuel utilization for 10 h, followed by measurement of the resulting carbon mass. The effect of current density was explored by measuring carbon mass under conditions known to be prone to coking while increasing the current density until the cell was carbon-free. The SC coking thresholds were measured to be ∼1.04 and ∼1.18 at 600 and 700 °C, respectively. Current density experiments validated the thresholds measured with respect to fuel utilization and steam:carbon ratio. Coking thresholds at 600 °C could be predicted with thermodynamic equilibrium calculations when the Gibbs free energy of carbon was appropriately modified. Here, the Gibbs free energy of carbon on nickel-based anode support cermets was measured to be -6.91 ± 0.08 kJ mol-1. The results of this two part publication show that thermodynamic equilibrium calculations with appropriate modification to the Gibbs free energy of solid-phase carbon can be used to predict coking thresholds on nickel-based anodes at 600-700 °C.

Numerical simulation of projectile impact on mild steel armour plates using LS-DYNA, Part II: Parametric studies

OpenAIRE

Raguraman, M; Deb, A; Gupta, NK; Kharat, DK

2008-01-01

In Part I of the current two-part series, a comprehensive simulation-based study of impact of Jacketed projectiles on mild steel armour plates has been presented. Using the modelling procedures developed in Part I, a number of parametric studies have been carried out for the same mild steel plates considered in Part I and reported here in Part II. The current investigation includes determination of ballistic limits of a given target plate for different projectile diameters and impact velociti...

Experience with lifetime limits for EBR-II core components

International Nuclear Information System (INIS)

Lambert, J.D.B.; Smith, R.N.; Golden, G.H.

1987-01-01

The Experimental Breeder Reactor No. 2 (EBR-II) is operated for the US Department of Energy by Argonne National Laboratory and is located on the Idaho National Engineering Laboratory where most types of American reactor were originally tested. EBR-II is a complete electricity-producing power plant now in its twenty-fourth year of successful operation. During this long history the reactor has had several concurrent missions, such as demonstration of a closed Liquid-Metal Reactor (LMR) fuel cycle (1964-69); as a steady-state irradiation facility for fuels and materials (1970 onwards); for investigating effects of operational transients on fuel elements (from 1981); for research into the inherent safety aspects of metal-fueled LMR's (from 1983); and, most recently, for demonstration of the Integral Fast Reactor (IFR) concept using U-Pu-Zr fuels. This paper describes experience gained at EBR-II in defining lifetime limits for LMR core components, particularly fuel elements

IFR fuel cycle--pyroprocess development

International Nuclear Information System (INIS)

Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

1992-01-01

The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

Energy Technology Data Exchange (ETDEWEB)

Steven R. Sherman

2005-04-01

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/or to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium

Fabrication of preliminary fuel rods for SFR

International Nuclear Information System (INIS)

Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

2012-01-01

Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

An alternative LEU design for the FRM-II

International Nuclear Information System (INIS)

Hanan, N.A.; Mo, S.C.; Smith, R.S.; Matos, J.E.

1996-01-01

The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm 3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance. LEU silicide fuel with 4.5 g/cm 3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. The following issues raised by TUM were addressed in Ref. 1: qualification of HEU and LEU silicide fuels, gamma heating in the heavy water reflector, radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. The conclusions of these analyses are summarized below. This paper addresses three additional safety issues that were raised by TUM in Ref. 2: stability of the involute fuel plates, a hypothetical accident involving the configuration of the reflector, and a loss of primary coolant flow transient due to an interrupted power supply. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility

Irradiation test and performance evaluation of DUPIC fuel

International Nuclear Information System (INIS)

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

2002-05-01

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

Methanol Fuel Cell

Science.gov (United States)

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.

The nuclear fuel cycle: (2) fuel element manufacture

International Nuclear Information System (INIS)

Doran, J.

1976-01-01

Large-scale production of nuclear fuel in the United Kingdom is carried out at Springfields Works of British Nuclear Fuels Ltd., a company formed from the United Kingdom Atomic Energy Authority in 1971. The paper describes in some detail the Springfields Works processes for the conversion of uranium ore concentrate to uranium tetrafluoride, then conversion of the tetrafluoride to either uranium metal for cladding in Magnox to form fuel for the British Mk I gas-cooled reactors, or to uranium hexafluoride for enrichment of the fissile 235 U isotope content at the Capenhurst Works of BNFL. Details are given of the reconversion at Springfields Works of this enriched uranium hexafluoride to uranium dioxide, which is pelleted and then clad in either stainless steel or zircaloy containers to form the fuel assemblies for the British Mk II AGR or advanced gas-cooled reactors or for the water reactor fuels. (author)

HERBICIDAS INIBIDORES DO FOTOSSISTEMA II – PARTE II / PHOTOSYSTEM II INHIBITOR HERBICIDES - PART

Directory of Open Access Journals (Sweden)

ILCA P. DE F. E SILVA

2013-11-01

Full Text Available Os herbicidas inibidores do fotossistema II (PSII ligam-se ao sítio da QB localizado na proteína D1 o qual se localiza na membrana dos tilacóides dos cloroplastos, causando, o bloqueia do transporte de elétrons da QA para QB, tendo como consequência, a peroxidação dos lipídios. Os principais fatores que afetam a evolução da resistência de plantas daninhas aos herbicidas têm sido agrupados em: genéticos, bioecológicos e agronômicos. A resistência de plantas daninhas a herbicidas é definida como a habilidade de uma planta sobreviver e reproduzir, após exposição a uma dose de herbicida normalmente letal para um biótipo normal da planta. A seletividade de um herbicida está relacionada à capacidade de eliminar plantas daninhas sem interferir na qualidade da planta de interesse econômico.

Spent fuel management newsletter. No. 2

International Nuclear Information System (INIS)

1993-04-01

This issue of the newsletter consists of two parts. The first part describes the IAEA Secretariat activities - work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes. The second part contains country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage and treatment of spent fuel

Spent fuel management newsletter. No. 2

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-04-01

This issue of the newsletter consists of two parts. The first part describes the IAEA Secretariat activities - work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes. The second part contains country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage and treatment of spent fuel.

Briquette fuel production from wastewater sludge of beer industry and biodiesel production wastes

Science.gov (United States)

Nusong, P.; Puajindanetr, S.

2018-04-01

The production of industrial wastes is increasing each year. Current methods of waste disposal are severely impacting the environment. Utilization of industrial wastes as an alternative material for fuel is gaining interest due to its environmental friendliness. Thus, the objective of this research was to study the optimum condition for fuel briquettes produced from wastewater sludge of the beer industry and biodiesel production wastes. This research is divided into two parts. Part I will study the effects of carbonization of brewery wastewater sludge for high fixed carbon. Part II will study the ratio between brewery wastewater sludge and bleaching earth for its high heating value. The results show that the maximum fixed carbon of 10.01% by weight was obtained at a temperature of 350 °C for 30 minutes. The appropriate ratio of brewery wastewater sludge and bleaching earth by weight was 95:5. This condition provided the highest heating value of approximately 3548.10 kcal/kg.

Polycystic ovary syndrome: a review for dermatologists: Part II. Treatment.

Science.gov (United States)

Buzney, Elizabeth; Sheu, Johanna; Buzney, Catherine; Reynolds, Rachel V

2014-11-01

Dermatologists are in a key position to treat the manifestations of polycystic ovary syndrome (PCOS). The management of PCOS should be tailored to each woman's specific goals, reproductive interests, and particular constellation of symptoms. Therefore, a multidisciplinary approach is recommended. In part II of this continuing medical education article, we present the available safety and efficacy data regarding treatments for women with acne, hirsutism, and androgenetic alopecia. Therapies discussed include lifestyle modification, topical therapies, combined oral contraceptives, antiandrogen agents, and insulin-sensitizing drugs. Treatment recommendations are made based on the current available evidence. Copyright © 2014 American Academy of Dermatology, Inc. Published by Elsevier Inc. All rights reserved.

Final project report, TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

International Nuclear Information System (INIS)

Montoya, G.M.

1992-01-01

This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project, including Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 m 3 of low-level solid radioactive waste and 5 m 3 of mixed waste

Recovery in soccer : part ii-recovery strategies.

Science.gov (United States)

Nédélec, Mathieu; McCall, Alan; Carling, Chris; Legall, Franck; Berthoin, Serge; Dupont, Gregory

2013-01-01

In the formerly published part I of this two-part review, we examined fatigue after soccer matchplay and recovery kinetics of physical performance, and cognitive, subjective and biological markers. To reduce the magnitude of fatigue and to accelerate the time to fully recover after completion, several recovery strategies are now used in professional soccer teams. During congested fixture schedules, recovery strategies are highly required to alleviate post-match fatigue, and then to regain performance faster and reduce the risk of injury. Fatigue following competition is multifactorial and mainly related to dehydration, glycogen depletion, muscle damage and mental fatigue. Recovery strategies should consequently be targeted against the major causes of fatigue. Strategies reviewed in part II of this article were nutritional intake, cold water immersion, sleeping, active recovery, stretching, compression garments, massage and electrical stimulation. Some strategies such as hydration, diet and sleep are effective in their ability to counteract the fatigue mechanisms. Providing milk drinks to players at the end of competition and a meal containing high-glycaemic index carbohydrate and protein within the hour following the match are effective in replenishing substrate stores and optimizing muscle-damage repair. Sleep is an essential part of recovery management. Sleep disturbance after a match is common and can negatively impact on the recovery process. Cold water immersion is effective during acute periods of match congestion in order to regain performance levels faster and repress the acute inflammatory process. Scientific evidence for other strategies reviewed in their ability to accelerate the return to the initial level of performance is still lacking. These include active recovery, stretching, compression garments, massage and electrical stimulation. While this does not mean that these strategies do not aid the recovery process, the protocols implemented up until

ANL calculational methodologies for determining spent nuclear fuel source term

International Nuclear Information System (INIS)

McKnight, R. D.

2000-01-01

Over the last decade Argonne National Laboratory has developed reactor depletion methods and models to determine radionuclide inventories of irradiated EBR-II fuels. Predicted masses based on these calculational methodologies have been validated using available data from destructive measurements--first from measurements of lead EBR-II experimental test assemblies and later using data obtained from processing irradiated EBR-II fuel assemblies in the Fuel Conditioning Facility. Details of these generic methodologies are described herein. Validation results demonstrate these methods meet the FCF operations and material control and accountancy requirements

Ocean Thermal Energy Converstion (OTEC) test facilities study program. Final report. Volume II. Part B

Energy Technology Data Exchange (ETDEWEB)

None

1977-01-17

Results are presented of an 8-month study to develop alternative non-site-specific OTEC facilities/platform requirements for an integrated OTEC test program which may include land and floating test facilities. Volume II--Appendixes is bound in three parts (A, B, and C) which together comprise a compendium of the most significant detailed data developed during the study. Part B provides an annotated test list and describes component tests and system tests.

Control structure design of a solid oxide fuel cell and a molten carbonate fuel cell integrated system: Top-down analysis

International Nuclear Information System (INIS)

Jienkulsawad, Prathak; Skogestad, Sigurd; Arpornwichanop, Amornchai

2017-01-01

Highlights: • Control structure of the combined fuel cell system is designed. • The design target is trade-off between power generation and carbon dioxide emission. • Constraints are considered according to fuel cell safe operation. • Eight variables have to be controlled to maximize profit. • Two control structures are purposed for three active constraint regions. - Abstract: The integrated system of a solid oxide fuel cell and molten carbonate fuel cell theoretically has very good potential for power generation with carbon dioxide utilization. However, the control strategy of such a system needs to be considered for efficient operation. In this paper, a control structure design for an integrated fuel cell system is performed based on economic optimization to select manipulated variables, controlled variables and control configurations. The objective (cost) function includes a carbon tax to get an optimal trade-off between power generation and carbon dioxide emission, and constraints include safe operation. This study focuses on the top-down economic analysis which is the first part of the design procedure. Three actively constrained regions as a function of the main disturbances, namely, the fuel and steam feed rates, are identified; each region represents different sets of active constraints. Under nominal operating conditions, the system operates in region I. However, operating the fuel cell system in region I and II can use the same structure, but in region III, a different control structure is required.

Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

International Nuclear Information System (INIS)

Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

2015-01-01

The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

Irradiation Test in HANARO of the Parts of an X-Gen Nuclear Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

2008-08-15

An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of an X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens requested by Westinghouse Co. and Hanyang university were also inserted. 389 KNF specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718 were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe (2 sets contain additional Nb-Ag) neutron fluence monitors installed in the capsule. The capsule was irradiated for 59.19days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 300 {approx} 420 .deg. C(for KNF specimens) up to a fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell.

Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

Energy Technology Data Exchange (ETDEWEB)

William J. Carmack

2012-05-01

Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

Not-for-profit versus for-profit health care providers--Part II: Comparing and contrasting their records.

Science.gov (United States)

Rotarius, Timothy; Trujillo, Antonio J; Liberman, Aaron; Ramirez, Bernardo

2006-01-01

The debate over which health care providers are most capably meeting their responsibilities in serving the public's interest continues unabated, and the comparisons of not-for-profit (NFP) versus for-profit (FP) hospitals remain at the epicenter of the discussion. From the perspective of available factual information, which of the two sides to this debate is correct? This article is part II of a 2-part series on comparing and contrasting the performance records of NFP health care providers with their FP counterparts. Although it is demonstrated that both NFP and FP providers perform virtuous and selfless feats on behalf of America's public, it is also shown that both camps have been accused of being involved in potentially willful clinical and administrative missteps. Part I provided the background information (eg, legal differences, perspectives on social responsibility, and types of questionable and fraudulent behavior) required to adequately understand the scope of the comparison issue. Part II offers actual comparisons of the 2 organizational structures using several disparate factors such as specific organizational behaviors, approach to the health care priorities of cost and quality, and business-focused goals of profits, efficiency, and community benefit.

Review of the IAEA nuclear fuel cycle and material section activities connected with nuclear fuel including WWER fuel

International Nuclear Information System (INIS)

Sokolov, F.

2001-01-01

Program activities on Nuclear Fuel Cycle and Materials cover the areas of: 1) raw materials (B.1.01); 2) fuel performance and technology (B.1.02); 3) pent fuel (B.1.03); 4) fuel cycle issues and information system (B.1.04); 5) support to technical cooperation activities (B.1.05). The IAEA activities in fuel performance and technology in 2001 include organization of the fuel experts meetings and completion of the Co-ordinate Research Projects (CRP). The special attention is given to the advanced post-irradiation examination techniques for water reactor fuel and fuel behavior under transients and LOCA conditions. An international research program on modeling of activity transfer in primary circuit of NPP is finalized in 2001. A new CRP on fuel modeling at extended burnup (FUMEX II) has planed to be carried out during the period 2002-2006. In the area of spent fuel management the implementation of burnup credit (BUC) in spent fuel management systems has motivated to be used in criticality safety applications, based on economic consideration. An overview of spent fuel storage policy accounting new fuel features as higher enrichment and final burnup, usage of MOX fuel and prolongation of the term of spent fuel storage is also given

The new German neutron source FRM-II

International Nuclear Information System (INIS)

Nuding, M.

2003-01-01

The 'Technische Universitaet Muenchen' has built a new high-flux research reactor, the 'Forschungsreaktor Muenchen'-II. This new reactor will replace the 'Forschungsreaktor Muenchen' which has been operated very successfully for about 43 years. The 'Forschungsreaktor Muenchen'-II has been developed with first priority for beam-tube experiments, but it will also provide possibilities for irradiation experiments or isotope production. The reactor was designed to obtain a high and spectrally pure thermal neutron flux is available in a large volume outside of the core, where it is accessible for experimental use. In addition to beam-tubes which will end in the thermal neutron field there will be beam-tubes that will provide - with the help of 'spectrum shifters' -cold; hot and fast neutrons. Even through the thermal power of the 'Forschungsreaktor Muenchen'-II was limited to 20 MW an unperturbed maximum thermal neutron flux of about 8 x 10 14 cm -2 s -1 will be reached. Because of its 'compact-core-concept' the 'Forschungsreaktor Muenchen'-II will have the best flux-to-power-ratio worldwide: The fuel element and its highly enriched U 3 Si 2 -Al-fuel were tested during the licensing procedure of the 'Forschungsreaktor Muenchen'-II. Within the scope of this 'hydraulic test' the stability and the vibration behavior of the fuel plates as well as the long-tem behavior of the fuel element were investigated (Authors)

Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part II

Energy Technology Data Exchange (ETDEWEB)

NONE

1978-07-01

The Summary Report - Part II of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - includes reports on development of fast reactors in France from 1977 to 1978; review of the activities related to fast reactors in Germany; status of fast breeder reactors development in Belgium and Netherlands; status of activities related to fast reactors in USSR, Japan USA, UK and Italy.

Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part II

International Nuclear Information System (INIS)

1978-07-01

The Summary Report - Part II of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - includes reports on development of fast reactors in France from 1977 to 1978; review of the activities related to fast reactors in Germany; status of fast breeder reactors development in Belgium and Netherlands; status of activities related to fast reactors in USSR, Japan USA, UK and Italy

A legacy of struggle: the OSHA ergonomics standard and beyond, Part II.

Science.gov (United States)

Delp, Linda; Mojtahedi, Zahra; Sheikh, Hina; Lemus, Jackie

2014-11-01

The OSHA ergonomics standard issued in 2000 was repealed within four months through a Congressional resolution that limits future ergonomics rulemaking. This section continues the conversation initiated in Part I, documenting a legacy of struggle for an ergonomics standard through the voices of eight labor, academic, and government key informants. Part I summarized important components of the standard; described the convergence of labor activism, research, and government action that laid the foundation for a standard; and highlighted the debates that characterized the rulemaking process. Part II explores the anti-regulatory political landscape of the 1990s, as well as the key opponents, power dynamics, and legal maneuvers that led to repeal of the standard. This section also describes the impact of the ergonomics struggle beyond the standard itself and ends with a discussion of creative state-level policy initiatives and coalition approaches to prevent work-related musculoskeletal disorders (WMSDs) in today's sociopolitical context.

Cyclopentane combustion. Part II. Ignition delay measurements and mechanism validation

KAUST Repository

Rachidi, Mariam El

2017-06-12

This study reports cyclopentane ignition delay measurements over a wide range of conditions. The measurements were obtained using two shock tubes and a rapid compression machine, and were used to test a detailed low- and high-temperature mechanism of cyclopentane oxidation that was presented in part I of this study (Al Rashidi et al., 2017). The ignition delay times of cyclopentane/air mixtures were measured over the temperature range of 650–1350K at pressures of 20 and 40atm and equivalence ratios of 0.5, 1.0 and 2.0. The ignition delay times simulated using the detailed chemical kinetic model of cyclopentane oxidation show very good agreement with the experimental measurements, as well as with the cyclopentane ignition and flame speed data available in the literature. The agreement is significantly improved compared to previous models developed and investigated at higher temperatures. Reaction path and sensitivity analyses were performed to provide insights into the ignition-controlling chemistry at low, intermediate and high temperatures. The results obtained in this study confirm that cycloalkanes are less reactive than their non-cyclic counterparts. Moreover, cyclopentane, a high octane number and high octane sensitivity fuel, exhibits minimal low-temperature chemistry and is considerably less reactive than cyclohexane. This study presents the first experimental low-temperature ignition delay data of cyclopentane, a potential fuel-blending component of particular interest due to its desirable antiknock characteristics.

Cyclopentane combustion. Part II. Ignition delay measurements and mechanism validation

KAUST Repository

Rachidi, Mariam El; Má rmol, Juan C.; Banyon, Colin; Sajid, Muhammad Bilal; Mehl, Marco; Pitz, William J.; Mohamed, Samah; Alfazazi, Adamu; Lu, Tianfeng; Curran, Henry J.; Farooq, Aamir; Sarathy, Mani

2017-01-01

This study reports cyclopentane ignition delay measurements over a wide range of conditions. The measurements were obtained using two shock tubes and a rapid compression machine, and were used to test a detailed low- and high-temperature mechanism of cyclopentane oxidation that was presented in part I of this study (Al Rashidi et al., 2017). The ignition delay times of cyclopentane/air mixtures were measured over the temperature range of 650–1350K at pressures of 20 and 40atm and equivalence ratios of 0.5, 1.0 and 2.0. The ignition delay times simulated using the detailed chemical kinetic model of cyclopentane oxidation show very good agreement with the experimental measurements, as well as with the cyclopentane ignition and flame speed data available in the literature. The agreement is significantly improved compared to previous models developed and investigated at higher temperatures. Reaction path and sensitivity analyses were performed to provide insights into the ignition-controlling chemistry at low, intermediate and high temperatures. The results obtained in this study confirm that cycloalkanes are less reactive than their non-cyclic counterparts. Moreover, cyclopentane, a high octane number and high octane sensitivity fuel, exhibits minimal low-temperature chemistry and is considerably less reactive than cyclohexane. This study presents the first experimental low-temperature ignition delay data of cyclopentane, a potential fuel-blending component of particular interest due to its desirable antiknock characteristics.

Fuel Behaviour Simulations in Fumex III CRP at NRI

International Nuclear Information System (INIS)

Valach, M.; Klouzal, J.; Dostal, M.; Zymak, J.

2013-01-01

NRI Rez plc took part in the previous coordinated research projects focused on fuel behaviour modelling held by the IAEA - FUMEX-I and FUMEX-II. These were very helpful for the development and validation of various codes used in the Nuclear Research Institute Rez (NRI) for the evaluation of the fuel rod thermomechanical behaviour. Based on the considerations of our needs related to the modeling for Czech NPPs we have performed basic parametric calculations of two LOCA cases (IFA-650.1 and IFA-650.2) and detailed evaluation WWER related cases Kola MIR ramp rods. The AREVA ''Idealized case'' and 16x16 LTA cases were also calculated because of the high burnup reached. Report summarises simulated cases in the frame of FUMEX III Project at the NRI Rez plc. (author)

EBR-II: summary of operating experience

International Nuclear Information System (INIS)

Perry, W.H.; Leman, J.D.; Lentz, G.L.; Longua, K.J.; Olson, W.H.; Shields, J.A.; Wolz, G.C.

1978-01-01

Experimental Breeder Reactor II (EBR-II) is an unmoderated, sodium-cooled reactor with a design power of 62.5 MWt. The primary cooling system is a submerged-pool type. The early operation of the reactor successfully demonstrated the feasibility of a sodium-cooled fast breeder reactor operating as an integrated reactor, power plant, and fuel-processing facility. In 1967, the role of EBR-II was reoriented from a demonstration plant to an irradiation facility. Many changes have been made and are continuing to be made to increase the usefulness of EBR-II for irradiation and safety tests. A review of EBR-II's operating history reveals a plant that has demonstrated high availability, stable and safe operating characteristics, and excellent performance of sodium components. Levels of radiation exposure to the operating and maintenance workers have been low; and fission-gas releases to the atmosphere have been minimal. Driver-fuel performance has been excellent. The repairability of radioactive sodium components has been successfully demonstrated a number of times. Recent highlights include installation and successful operation of (1) the hydrogen-meter leak detectors for the steam generators, (2) the cover-gas-cleanup system and (3) the cesium trap in the primary sodium. Irradiations now being conducted in EBR-II include the run-beyond-cladding breach fuel tests for mixed-oxide and carbide elements. Studies are in progress to determine EBR-II's capability for conducting important ''operational safety'' tests. These tests would extend the need and usefulness of EBR-II into the 1980's

Title II, Part A: Don't Scrap It, Don't Dilute It, Fix It

Science.gov (United States)

Coggshall, Jane G.

2015-01-01

The Issue: Washington is taking a close look at Title II, Part A (Title IIA) of the Elementary and Secondary Education Act (ESEA) as Congress debates reauthorization. The program sends roughly $2.5 billion a year to all states and nearly all districts to "(1) increase student academic achievement through strategies such as improving teacher…

Instructional Climates in Preschool Children Who Are At-Risk. Part II: Perceived Physical Competence

Science.gov (United States)

Robinson, Leah E.; Rudisill, Mary E.; Goodway, Jacqueline D.

2009-01-01

In Part II of this study, we examined the effect of two 9-week instructional climates (low-autonomy [LA] and mastery motivational climate [MMC]) on perceived physical competence (PPC) in preschoolers (N = 117). Participants were randomly assigned to an LA, MMC, or comparison group. PPC was assessed by a pretest, posttest, and retention test with…

The history, genotoxicity, and carcinogenicity of carbon-based fuels and their emissions. Part 2: solid fuels.

Science.gov (United States)

Claxton, Larry D

2014-01-01

The combustion of solid fuels (like wood, animal dung, and coal) usually involves elevated temperatures and altered pressures and genotoxicants (e.g., PAHs) are likely to form. These substances are carcinogenic in experimental animals, and epidemiological studies implicate these fuels (especially their emissions) as carcinogens in man. Globally, ∼50% of all households and ∼90% of all rural households use solid fuels for cooking or heating and these fuels often are burnt in simple stoves with very incomplete combustion. Exposed women and children often exhibit low birth weight, increased infant and perinatal mortality, head and neck cancer, and lung cancer although few studies have measured exposure directly. Today, households that cannot meet the expense of fuels like kerosene, liquefied petroleum gas, and electricity resort to collecting wood, agricultural residue, and animal dung to use as household fuels. In the more developed countries, solid fuels are often used for electric power generation providing more than half of the electricity generated in the United States. The world's coal reserves, which equal approximately one exagram, equal ∼1 trillion barrels of crude oil (comparable to all the world's known oil reserves) and could last for 600 years. Studies show that the PAHs that are identified in solid fuel emissions react with NO2 to form direct-acting mutagens. In summary, many of the measured genotoxicants found in both the indoor and electricity-generating combustors are the same; therefore, the severity of the health effects vary with exposure and with the health status of the exposed population. Copyright © 2014. Published by Elsevier B.V.

A natural-gas fuel processor for a residential fuel cell system

Science.gov (United States)

Adachi, H.; Ahmed, S.; Lee, S. H. D.; Papadias, D.; Ahluwalia, R. K.; Bendert, J. C.; Kanner, S. A.; Yamazaki, Y.

A system model was used to develop an autothermal reforming fuel processor to meet the targets of 80% efficiency (higher heating value) and start-up energy consumption of less than 500 kJ when operated as part of a 1-kWe natural-gas fueled fuel cell system for cogeneration of heat and power. The key catalytic reactors of the fuel processor - namely the autothermal reformer, a two-stage water gas shift reactor and a preferential oxidation reactor - were configured and tested in a breadboard apparatus. Experimental results demonstrated a reformate containing ∼48% hydrogen (on a dry basis and with pure methane as fuel) and less than 5 ppm CO. The effects of steam-to-carbon and part load operations were explored.

Posttest examination results of recent treat tests on metal fuel

International Nuclear Information System (INIS)

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

1986-01-01

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

Optimal recombination in genetic algorithms for combinatorial optimization problems: Part II

Directory of Open Access Journals (Sweden)

Eremeev Anton V.

2014-01-01

Full Text Available This paper surveys results on complexity of the optimal recombination problem (ORP, which consists in finding the best possible offspring as a result of a recombination operator in a genetic algorithm, given two parent solutions. In Part II, we consider the computational complexity of ORPs arising in genetic algorithms for problems on permutations: the Travelling Salesman Problem, the Shortest Hamilton Path Problem and the Makespan Minimization on Single Machine and some other related problems. The analysis indicates that the corresponding ORPs are NP-hard, but solvable by faster algorithms, compared to the problems they are derived from.

An approach to WWER fuels with BaCo

International Nuclear Information System (INIS)

Marino, A.; Demarco, G.

2008-01-01

BaCo is a code for the simulation of the behaviour of a nuclear fuel rod under operation conditions. BaCo, a quasi 2D code based on a finite differences scheme, has been used for simulating PHWR, CANDU, PWR, BWR, MOX, WWER, and experimental fuel rods. We improve the performance of BaCo with a set of tools based on the method of finite elements for 3D analysis of the stress-strain state. We can simulate any UO 2 pellet geometry. Standard WWER-440 fuel assemblies irradiated in the Kola-3 reactor of the CRP FUMEX II of the IAEA were the first WWER simulations with BaCo. We find a very good agreement among our calculations, the experimental results and other qualified fuel codes. We present the BaCo code and our results for PWR and WWER fuels of the CRP FUMEX II, the 3D analysis of WWER fuel pellet and the projections of these results with the Argentinean nuclear fuels development. (authors)

The basic science of dermal fillers: past and present Part II: adverse effects.

Science.gov (United States)

Gilbert, Erin; Hui, Andrea; Meehan, Shane; Waldorf, Heidi A

2012-09-01

The ideal dermal filler should offer long-lasting aesthetic improvement with a minimal side-effect profile. It should be biocompatible and stable within the injection site, with the risk of only transient undesirable effects from injection alone. However, all dermal fillers can induce serious and potentially long-lasting adverse effects. In Part II of this paper, we review the most common adverse effects related to dermal filler use.

Biology and Mechanics of Blood Flows Part II: Mechanics and Medical Aspects

CERN Document Server

Thiriet, Marc

2008-01-01

Biology and Mechanics of Blood Flows presents the basic knowledge and state-of-the-art techniques necessary to carry out investigations of the cardiovascular system using modeling and simulation. Part II of this two-volume sequence, Mechanics and Medical Aspects, refers to the extraction of input data at the macroscopic scale for modeling the cardiovascular system, and complements Part I, which focuses on nanoscopic and microscopic components and processes. This volume contains chapters on anatomy, physiology, continuum mechanics, as well as pathological changes in the vasculature walls including the heart and their treatments. Methods of numerical simulations are given and illustrated in particular by application to wall diseases. This authoritative book will appeal to any biologist, chemist, physicist, or applied mathematician interested in the functioning of the cardiovascular system.

Formulation, computation and improvement of steady state security margins in power systems. Part II: Results

International Nuclear Information System (INIS)

Echavarren, F.M.; Lobato, E.; Rouco, L.; Gomez, T.

2011-01-01

A steady state security margin for a particular operating point can be defined as the distance from this initial point to the secure operating limits of the system. Four of the most used steady state security margins are the power flow feasibility margin, the contingency feasibility margin, the load margin to voltage collapse, and the total transfer capability between system areas. This is the second part of a two part paper. Part I has proposed a novel framework of a general model able to formulate, compute and improve any steady state security margin. In Part II the performance of the general model is validated by solving a variety of practical situations in modern real power systems. Actual examples of the Spanish power system will be used for this purpose. The same computation and improvement algorithms outlined in Part I have been applied for the four security margins considered in the study, outlining the convenience of defining a general framework valid for the four of them. The general model is used here in Part II to compute and improve: (a) the power flow feasibility margin (assessing the influence of the reactive power generation limits in the Spanish power system), (b) the contingency feasibility margin (assessing the influence of transmission and generation capacity in maintaining a correct voltage profile), (c) the load margin to voltage collapse (assessing the location and quantity of loads that must be shed in order to be far away from voltage collapse) and (d) the total transfer capability (assessing the export import pattern of electric power between different areas of the Spanish system). (author)

Formulation, computation and improvement of steady state security margins in power systems. Part II: Results

Energy Technology Data Exchange (ETDEWEB)

Echavarren, F.M.; Lobato, E.; Rouco, L.; Gomez, T. [School of Engineering of Universidad Pontificia Comillas, C/Alberto Aguilera, 23, 28015 Madrid (Spain)

2011-02-15

A steady state security margin for a particular operating point can be defined as the distance from this initial point to the secure operating limits of the system. Four of the most used steady state security margins are the power flow feasibility margin, the contingency feasibility margin, the load margin to voltage collapse, and the total transfer capability between system areas. This is the second part of a two part paper. Part I has proposed a novel framework of a general model able to formulate, compute and improve any steady state security margin. In Part II the performance of the general model is validated by solving a variety of practical situations in modern real power systems. Actual examples of the Spanish power system will be used for this purpose. The same computation and improvement algorithms outlined in Part I have been applied for the four security margins considered in the study, outlining the convenience of defining a general framework valid for the four of them. The general model is used here in Part II to compute and improve: (a) the power flow feasibility margin (assessing the influence of the reactive power generation limits in the Spanish power system), (b) the contingency feasibility margin (assessing the influence of transmission and generation capacity in maintaining a correct voltage profile), (c) the load margin to voltage collapse (assessing the location and quantity of loads that must be shed in order to be far away from voltage collapse) and (d) the total transfer capability (assessing the export import pattern of electric power between different areas of the Spanish system). (author)

Factors controlling metal fuel lifetime

International Nuclear Information System (INIS)

Porter, D.L.; Hofman, G.L.; Seidel, B.R.; Walters, L.C.

1986-01-01

The reliability of metal fuel elements is determined by a fuel burnup at which a statistically predicted number of fuel breaches would occur, the number of breaches determined by the amount of free fission gas which a particular reactor design can tolerate. The reliability is therefore measured using experimentally determined breach statistics, or by modelling fuel element behavior and those factors which contribute to cladding breach. The factors are fuel/cladding mechanical and chemical interactions, fission gas pressure, fuel phase transformations involving volume changes, and fission product effects on cladding integrity. Experimental data for EBR-II fuel elements has shown that the primary, and perhaps the only significant factor affecting metal fuel reliability, is the pressure-induced stresses caused by fission gas release. Other metal fuel/cladding systems may perform similarly

Failed fuel identification techniques for liquid-metal cooled reactors

International Nuclear Information System (INIS)

Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

1995-01-01

The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures

Design report for an annular fuel element for accommodation of a carbide test bundle on the ring position of the KNK II/2 test zone

International Nuclear Information System (INIS)

Haefner, H.E.

1982-03-01

This report describes an annular oxide element with Mark II rods for accommodation of a 19-pin carbide test bundle on position 201 in the test zone of the second core of KNK II as well as its behavior during the period of operation. The ring element comprises within a driver wrapper in three rows of pins 102 fuel pins of 7.6 mm diameter and six structural rods for fixing the spark eroded spacers. The report deals with the ring element with its individual components fuel rod, bundle, wrappers, head and foot and describes methods, criteria and results concerning the design. The carbide test bundle to be accommodated by the annular carrier element will be treated in a separate report. The loadability of the annular element with its components is demonstrated by generally valid standards for strength criteria

Exploring Water Pollution. Part II

Science.gov (United States)

Rillo, Thomas J.

1975-01-01

This is part two of a three part article related to the science activity of exploring environmental problems. Part one dealt with background information for the classroom teacher. Presented here is a suggested lesson plan on water pollution. Objectives, important concepts and instructional procedures are suggested. (EB)

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

International Nuclear Information System (INIS)

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

1977-01-01

The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)

Part I: $\\beta$-delayed fission, laser spectroscopy and shape-coexistence studies with astatine beams; Part II: Delineating the island of deformation in the light gold isotopes by means of laser spectroscopy

CERN Document Server

Andreyev, Andrei

2013-01-01

Part I: $\\beta$-delayed fission, laser spectroscopy and shape-coexistence studies with astatine beams; Part II: Delineating the island of deformation in the light gold isotopes by means of laser spectroscopy

Final environmental statement. Final addendum to Part II: Manufacture of floating nuclear power plants by Offshore Power Systems. DOCKET-STN--50-437

International Nuclear Information System (INIS)

1978-06-01

This Addendum to Part II of the Final Environmental Statement related to manufacture of floating nuclear power plants by Offshore Power Systems (OPS), NUREG-0056, issued September 1976, was prepared by the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation. The staff's basic evaluation is presented in NUREG-0056. The current Addendum provides further consideration of a number of topics discussed in NUREG-0056, particularly additional consideration of shore zone siting at estuarine and ocean regions. This Summary and Conclusions recapitulates and is cumulative for Part II of the FES and the current Addendum. Augmentations to the Summary and Conclusions presented in Part II of the FES and arising from the evaluations contained in this Addendum are italicized

Stationary liquid fuel fast reactor SLFFR – Part I: Core design

Energy Technology Data Exchange (ETDEWEB)

Jing, T.; Yang, G.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

2016-12-15

Highlights: • An innovative fast reactor concept SLFFR based on liquid metal fuel is proposed for TRU burning. • A compact core design of 1000 MWt SLFFR is developed to achieve a zero conversion ratio and passive safety. • The core size and the control requirement are significantly reduced compared to the conventional solid fuel reactor with same conversion ratio. - Abstract: For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative.

Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

1978-12-01

The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

Storage arrangement for nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Wade, E.E.

1977-01-01

Said invention is intended for providing an arrangement of spent fuel assembly storage inside which the space is efficiently used without accumulating a critical mass. The storage is provided for long fuel assemblies having along their longitudinal axis an active part containing the fuel and an inactive part empty of fuel. Said storage arrangement comprises a framework constituting some long-shaped cells designed so as each of them can receive a fuel assembly. Means of axial positioning of said assembly in a cell make it possible to support the fuel assemblies inside the framework according to a spacing ratio, along the cell axis, such as the active part of an assembly is adjacent to the inactive part of the adjacent assemblies [fr

Current antiviral drugs and their analysis in biological materials - Part II: Antivirals against hepatitis and HIV viruses.

Science.gov (United States)

Nováková, Lucie; Pavlík, Jakub; Chrenková, Lucia; Martinec, Ondřej; Červený, Lukáš

2018-01-05

This review is a Part II of the series aiming to provide comprehensive overview of currently used antiviral drugs and to show modern approaches to their analysis. While in the Part I antivirals against herpes viruses and antivirals against respiratory viruses were addressed, this part concerns antivirals against hepatitis viruses (B and C) and human immunodeficiency virus (HIV). Many novel antivirals against hepatitis C virus (HCV) and HIV have been introduced into the clinical practice over the last decade. The recent broadening portfolio of these groups of antivirals is reflected in increasing number of developed analytical methods required to meet the needs of clinical terrain. Part II summarizes the mechanisms of action of antivirals against hepatitis B virus (HBV), HCV, and HIV, their use in clinical practice, and analytical methods for individual classes. It also provides expert opinion on state of art in the field of bioanalysis of these drugs. Analytical methods reflect novelty of these chemical structures and use by far the most current approaches, such as simple and high-throughput sample preparation and fast separation, often by means of UHPLC-MS/MS. Proper method validation based on requirements of bioanalytical guidelines is an inherent part of the developed methods. Copyright © 2017 Elsevier B.V. All rights reserved.

Transport of nuclear material (Part II)

International Nuclear Information System (INIS)

Staake, Theo; Schmidt, Thomas

1983-01-01

Providing a complete back-end service for MTR reactors is one of the fundamental and traditional tasks of TRANSNUKLEAR GmbH (TN). TN's services in this field cover everything from supplying the ideal transport cask, providing technical assistance during the loading operation, obtaining the necessary package approval and transport licenses, providing the required insurance cover, carrying out the transport, right thru to settling the reprocessing contract. Up until 1976, TN carried out transports of MTR fuel elements to the European reprocessing plants at Mol in Belgium and Marcoule in France. In all, some 1000 fuel elements were transported in this p e ri od. However, following the decision by these plants not to reprocess these elements anymore, subsequent transports had to be made to the US-DOE reprocessing plants. TN pooled together the interests of all her MTR customers and signed a reprocessing contract with the US-DOE, which ensured a complete back-end service for these reactors well into the future. In close cooperation with our associated company, Transnuclear Inc. in New York and Washington, a new transport concept was developed, which proved itself to be both economic and reliable. Up to now, a total of about 2050 MTR fuel elements have been transported by TN-Germany to the USA in 65 separate shipments. The total number of shipments performed by the TN group is 165 shipments. All shipments were carried out routinely without any incident. In March this year, the US-DOE made use of a clause in the contract, in which 90 days' notice was given of a change in reprocessing plant. Whereas previously all elements had been taken to the Savannah River Plant (SRP) in South Carolina, in future all elements have to go to the Idaho Chemical Processing Plant (ICPP) near Idaho Falls. This change presented TN with the not inconsiderable problem of finding a suitable transport route. Due to the large number of influencing factors, the TN-group carried out a special

Nuclear reactor fuel element with a cluster of parallel fuel pins

International Nuclear Information System (INIS)

Macfall, D.; Butterfield, C.E.; Butterfield, R.S.

1977-01-01

An improvement of the design of nuclear reactor fuel elements is described and illustrated by the example of a gas-cooled, graphite-moderated nuclear reactor. The fuel element has a cluster of parallel fuel pins with an outer can of structure material and an inner sleeve, as well as tie bars and spacing devices for all of these parts. The fuel element designed according to the invention allows lasy assembling and disassembling before and after use. During use, no relative axial motions are possible; nevertheless, the graphite sleeve is at no time subject to tensile stress: the individual parts are held in position from below by a single holding device. (UWI) [de

Nuclear power fuel cycle

International Nuclear Information System (INIS)

Havelka, S.; Jakesova, L.

1982-01-01

Economic problems are discussed of the fuel cycle (cost of the individual parts of the fuel cycle and the share of the fuel cycle in the price of 1 kWh), the technological problems of the fuel cycle (uranium ore mining and processing, uranium isotope enrichment, the manufacture of fuel elements, the building of long-term storage sites for spent fuel, spent fuel reprocessing, liquid and gaseous waste processing), and the ecologic aspects of the fuel cycle. (H.S.)

Advances in AGR fuel fabrication - now and the future

International Nuclear Information System (INIS)

Bleasdale, P.A.

1995-01-01

To date, over 3 million AGR fuel pins have been manufactured at Springfields for the UK AGR programme. During this time, AGR fuel design and manufacture has developed and evolved in response to the needs of the reactor operators to enhance fuel reliability and performance. More recently, major advances have been made in the systems and organisational culture which support fuel manufacture at Fuel Division. The introduction of MRP II in 1989 into Fuel Division enabled significant reductions in stock and work-in-progress, together with reductions in manufacturing lead times. Other successful initiatives introduced into Fuel Division have been Just-in-Time (JIT) and AST (Additional Skills Training) which have built on the success of MRP II. All of these initiatives are evidence of Fuel Division's ''Total Quality'' approach to fabricating fuel. Fuel Division is currently in the final stages of commissioning the New Oxide Fuels Complex (NOFC) where both AGR and PWR fuel will be manufactured to the highest standards of quality, safety and environmental protection. NOFC is a totally integrated plant which represents a Pound 200M investment, demonstrating Fuel Division's commitment to building on its 40+ years of fuel fabrication experience and ensuring secure supply of fuel to its customers for years to come. (author)