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

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 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

A user input manual for single fuel rod behaviour analysis code FEMAXI-III

International Nuclear Information System (INIS)

Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

1983-03-01

Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [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, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

2014-02-15

Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

Warship Radar Signatures (Ship Survivability Part III-A)

NARCIS (Netherlands)

Galle, L.F.; Heemskerk, H.J.M.; Ewijk, L.J. van

2000-01-01

Radar Cross Section (RCS) management is of paramount importance for a warships's survivability. In this first part of the paper (Part III-A), the operational benefits of low RCS will be explained. Basic RCS theory, measurement and simulation techniques will be addressed. The RCS of representative

Performance limits of coated particle fuel. Part III. Fission product migration in HTR fuel

Energy Technology Data Exchange (ETDEWEB)

Nabielek, H.; Hick, H.; Wagner-Loffler, M.; Voice, E. H.

1974-06-15

A general introduction and literature survey to the physics and mathematics of fission product migration in HTR fuel is given as well as a review of available experimental results and their evaluation in terms of models and materials data.

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

12 CFR Appendix III to Part 27 - Fair Housing Lending Inquiry/Application Log Sheet

Science.gov (United States)

2010-01-01

... 12 Banks and Banking 1 2010-01-01 2010-01-01 false Fair Housing Lending Inquiry/Application Log Sheet III Appendix III to Part 27 Banks and Banking COMPTROLLER OF THE CURRENCY, DEPARTMENT OF THE TREASURY FAIR HOUSING HOME LOAN DATA SYSTEM Pt. 27, App. III Appendix III to Part 27—Fair Housing Lending...

Solar neutrino oscillation parameters after SNO Phase-III and SAGE Part-III

International Nuclear Information System (INIS)

Yang Ping; Liu Qiuyu

2009-01-01

We analyse the recently published results from solar neutrino experiments SNO Phase-III and SAGE Part-III and show their constraints on solar neutrino oscillation parameters, especially for the mixing angle θ 12 . Through a global analysis using all existing data from SK, SNO, Ga and Cl radiochemical experiments and long base line reactor experiment KamLAND , we obtain the parameters Δm 12 2 =7.684 -0.208 +0.212 x 10 -5 eV 2 , tan 2 θ 12 =0.440 -0.057 +0.059 . We also find that the discrepancy between the KamLAND and solar neutrino results can be reduced by choosing a small non-zero value for the mixing angle θ 13 . (authors)

Innovative nuclear fuels and applications. Part 1: limits of today's fuels and concepts for innovative fuels. Part 2: materials properties, irradiation performance and gaps in our knowledge

International Nuclear Information System (INIS)

Matzke, H.

2000-01-01

Part I of this contribution on innovative nuclear fuels gives a summary of current developments and problems of today's fuels, i.e. enriched UO 2 and UO 2 with a few % of PUO 2 (MOX fuel) or Gd 2 O 3 (as burnable neutron poison). The problems and property changes caused by high burnups (e.g. degradation of the thermal conductivity, polygonization or formation of the rim-structure) are discussed. Subsequently, the concepts for new fuels to burn excess Pu and to achieve an effective transmutation of the minor actinides Np, Am and Cm are treated. The criteria for the choice of suitable fuels and different fuel types (high Pu-content fuels, nitrides, U-free fuels, inert matrix supported fuels, cercers, cermets, etc.) are discussed. Part II of this contribution on innovative nuclear fuels deals with the properties of relevance of the different materials suggested to be used in innovative fuels which range from pure actinide fuel such as PuN and AmO 2 to spinel MgAl 2 O 4 and zircon ZrSiO 4 for inert matrix-based fuels, etc. The available knowledge on materials research aspects is summarized with emphasis on the physics of radiation damage. It is shown that significant gaps in the present knowledge exist, e.g. for the minor actinide compounds, and suggestions are made to fill these gaps in order to achieve a sufficient data base to design and operate suitable innovative fuels in a near future. (author)

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.

A novel concept of QUADRISO particles Part III: applications to the plutonium-thorium fuel cycle

International Nuclear Information System (INIS)

Talamo, A.

2009-01-01

In the present study, a plutonium-thorium fuel cycle is investigated including the 233 U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing 233 U. The above configuration has been compared with a configuration where fissile (neptunium-plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the 233 U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.

10 CFR Appendix II to Part 504 - Fuel Price Computation

Science.gov (United States)

2010-01-01

... DEPARTMENT OF ENERGY (CONTINUED) ALTERNATE FUELS EXISTING POWERPLANTS Pt. 504, App. II Appendix II to Part... effects of future real price increases for each fuel. The delivered price of an alternate fuel used to calculate delivered fuel expenses must reflect the petitioner's delivered price of the alternate fuel and...

18 CFR 410.1 - Basin regulations-Water Code and Administrative Manual-Part III Water Quality Regulations.

Science.gov (United States)

2010-04-01

... Code and Administrative Manual-Part III Water Quality Regulations. 410.1 Section 410.1 Conservation of... CODE AND ADMINISTRATIVE MANUAL-PART III WATER QUALITY REGULATIONS § 410.1 Basin regulations—Water Code and Administrative Manual—Part III Water Quality Regulations. (a) The Water Code of the Delaware River...

FEMAXI-III, a computer code for fuel rod performance analysis

International Nuclear Information System (INIS)

Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

1983-01-01

This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

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

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...

Standards in neurosonology. Part III

Directory of Open Access Journals (Sweden)

Joanna Wojczal

2016-06-01

Full Text Available The paper presents standards related to ultrasound imaging of the cerebral vasculature and structures. The aim of this paper is to standardize both the performance and description of ultrasound imaging of the extracranial and intracranial cerebral arteries as well as a study of a specific brain structure, i.e. substantia nigra hyperechogenicity. The following aspects are included in the description of standards for each ultrasonographic method: equipment requirements, patient preparation, study technique and documentation as well as the required elements of ultrasound description. Practical criteria for the diagnosis of certain pathologies in accordance with the latest literature were also presented. Furthermore, additional comments were included in some of the sections. Part I discusses standards for the performance, documentation and description of different ultrasound methods (Duplex, Doppler. Part II and III are devoted to standards for specific clinical situations (vasospasm, monitoring after the acute stage of stroke, detection of a right-to-left shunts, confirmation of the arrest of the cerebral circulation, an assessment of the functional efficiency of circle of Willis, an assessment of the cerebrovascular vasomotor reserve as well as the measurement of substantia nigra hyperechogenicity.

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...

Eighth annual meeting, Vienna, Austria, 15-18 April 1975. Summary report. Part III

International Nuclear Information System (INIS)

1976-01-01

The Summary Report of the Eighth Annual Meeting of the International Working Group on Fast Reactors contains the minutes of the meeting (Part 1); papers which review the national programmes in the field of LMFBRs (Part 2) and the discussions on the review of national programmes (Part 3). The agenda of the meeting involved design, construction, operating experiences of demonstration fast power reactors, reprocessing of spent fuel from LMFBRs, reliability of decay heat removal systems, fuel failure mechanisms, safety of LMFBRs

Eighth annual meeting, Vienna, Austria, 15-18 April 1975. Summary report. Part III

Energy Technology Data Exchange (ETDEWEB)

NONE

1976-01-01

The Summary Report of the Eighth Annual Meeting of the International Working Group on Fast Reactors contains the minutes of the meeting (Part 1); papers which review the national programmes in the field of LMFBRs (Part 2) and the discussions on the review of national programmes (Part 3). The agenda of the meeting involved design, construction, operating experiences of demonstration fast power reactors, reprocessing of spent fuel from LMFBRs, reliability of decay heat removal systems, fuel failure mechanisms, safety of LMFBRs.

40 CFR Appendix II to Part 600 - Sample Fuel Economy Calculations

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Sample Fuel Economy Calculations II... FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Pt. 600, App. II Appendix II to Part 600—Sample Fuel Economy Calculations (a) This sample fuel economy calculation is applicable to...

Some Aspects of Facial Nerve Paralysis. Part III. Complications ...

African Journals Online (AJOL)

Some Aspects of Facial Nerve Paralysis. Part III. Complications, Prognosis and management. ... It should be possible to set a definite prognosis within 2 weeks after the onset of facial paralysis, and in many cases even sooner. In the prognosis of facial paralysis the aetiological and time factors involved, the completeness of ...

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part III

International Nuclear Information System (INIS)

2011-01-01

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

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...

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

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

Recent operational history of the new Sandia Pulsed Reactor III (SPR III)

International Nuclear Information System (INIS)

Schmidt, T.R.; Estes, B.F.; Reuscher, J.A.

1977-01-01

The Sandia Pulsed Reactor III (SPR III) is a fast-pulse research reactor which was designed and built at Sandia Laboratories and achieved criticality in August 1975. The reactor is now characterized and is in an operational configuration. The core consists of 18 fuel plates (258 kg fuel mass) of fully enriched uranium alloyed with 10 wt.% molybdenum. It is arranged in an annular configuration with an inside diameter of 17.78 cm, an outside diameter of 29.72 cm, and a height of 35.9 cm. The reactor core uses reflectors of copper and aluminum for control and an external bolting arrangement to secure the fuel plates. SPR III and SPR II are operated on an interchangeable basis using the same facility and control system. As of June 1977, SPR III has had over 240 operations with core temperatures up to 541 0 C

COMETHE III J a computer code for predicting mechanical and thermal behaviour of a fuel pin

International Nuclear Information System (INIS)

Verbeek, P.; Hoppe, N.

1976-01-01

The design of fuel pins for power reactors requires a realistic evaluation of their thermal and mechanical performances throughout their irradiation life. This evaluation involves the knowledge of a number of parameters, very intricate and interconnected, for example, the temperature, the restructuring and the swelling rates of the fuel pellets, the dimensions, the stresses and the strains in the clad, the composition and the properties of gases, the inner gas pressure etc. This complex problem can only be properly handled by a computer programme which analyses the fuel pin thermal and mechanical behaviour at successive steps of its irradiation life. This report presents an overall description of the COMETHE III-J computer programme, designed to calculate the integral performance of oxide fuel pins with cylindrical metallic cladding irradiated in thermal or fast flux. (author)

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...

Hanford spent nuclear fuel project recommended path forward, volume III: Alternatives and path forward evaluation supporting documentation

International Nuclear Information System (INIS)

Fulton, J.C.

1994-10-01

Volume I of the Hanford Spent Nuclear Fuel Project - Recommended Path Forward constitutes an aggressive series of projects to construct and operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. Volume II provided a comparative evaluation of four Alternatives for the Path Forward and an evaluation for the Recommended Path Forward. Although Volume II contained extensive appendices, six supporting documents have been compiled in Volume III to provide additional background for Volume II

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.

Regulations for safe transport of spent fuels from nuclear power plants in CMEA member countries. Part III

International Nuclear Information System (INIS)

Zizka, B.

1978-11-01

The regulations for safe transport of spent fuel from nuclear power plants in the CMEA member countries consist of general provisions, technical requirements for spent fuel transport, transport conditions, procedures for submitting reports on transport, regulations for transport and protection of radioactive material to be transported, procedures for customs clearance, technical and organizational measures for the prevention of hypothetical accidents and the elimination of their consequences. The bodies responsible for spent fuel transport in the CMEA member countries are listed. (J.B.)

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.

A structural modification of the two dimensional fuel behaviour analysis code FEMAXI-III with high-speed vectorized operation

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Ishiguro, Misako; Yamazaki, Takashi; Tokunaga, Yasuo.

1985-02-01

Though the two-dimensional fuel behaviour analysis code FEMAXI-III has been developed by JAERI in form of optimized scalar computer code, the call for more efficient code usage generally arized from the recent trends like high burn-up and load follow operation asks the code into further modification stage. A principal aim of the modification is to transform the already implemented scalar type subroutines into vectorized forms to make the programme structure efficiently run on high-speed vector computers. The effort of such structural modification has been finished on a fair way to success. The benchmarking two tests subsequently performed to examine the effect of the modification led us the following concluding remarks: (1) In the first benchmark test, comparatively high-burned three fuel rods that have been irradiated in HBWR, BWR, and PWR condition are prepared. With respect to all cases, a net computing time consumed in the vectorized FEMAXI is approximately 50 % less than that consumed in the original one. (2) In the second benchmark test, a total of 26 PWR fuel rods that have been irradiated in the burn-up ranges of 13-30 MWd/kgU and subsequently power ramped in R2 reactor, Sweden is prepared. In this case the code is purposed to be used for making an envelop of PCI-failure threshold through 26 times code runs. Before coming to the same conclusion, the vectorized FEMAXI-III consumed a net computing time 18 min., while the original FEMAXI-III consumed a computing time 36 min. respectively. (3) The effects obtained from such structural modification are found to be significantly attributed to saving a net computing time in a mechanical calculation in the vectorized FEMAXI-III code. (author)

Startup of Torrey Pines Mark III and Puerto Rico Nuclear Center reactors with TRIGA-FLIP fuel

Energy Technology Data Exchange (ETDEWEB)

Chesworth, R. H. [Gulf E and ES, San Diego, CA (United States)

1972-07-01

This paper discusses the characteristics of TRIGA FLIP cores in two different geometries: the normal TRIGA single-rod geometry as typified by the installation in the Torrey Pines Mark III reactor; and the four-rod cluster geometry as typified by the conversion core installed in the Puerto Rico Nuclear Center reactor at Mayaguez. In both reactors the fuel is 8-1/2 wt % uranium, 70% enriched in U-235. The hydrogen to zirconium atom ratio is 1.5 to 1.65 and the cladding material is stainless steel. The basic neutronic characteristics of the fuel in both reactor installations are briefly discussed.

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

The priority cases of the FUMEX-III exercises simulated with the TRANSURANUS code

International Nuclear Information System (INIS)

Boneva, S.

2011-01-01

The FUMEX-III project provides a good basis for testing common code priorities and the needs for further developments. The GAIN experiment contains results on four Gd 2 O 3 doped UO 2 rods and offers good opportunities for testing of the fuel performance codes in the case of Gd-doped fuel. A good agreement between the TRANSURANUS calculations and the measurements is achieved for the fuel and the cladding deformation. The FUMEX-III priority cases cover two rods from the GINNA reactor experiment: rod2 with fuel solid pellets, and rod4 with annular pellets and standard Zircaloy-4 cladding. Both rods were irradiated 5 cycles up to 52MWd/kgU. The simulations of the GINNA and US PWR experiments are part of the ongoing validation of the TRANSURANUS code - for different pellet design. The simulations of irradiation transients reveal the need for improving the fission gas release model, including burst release and release from the high burn-up structure

VTT ENIGMA Calculations for FUMEX-III CRP

International Nuclear Information System (INIS)

Tulkki, Ville

2013-01-01

International Atomic Energy Agency IAEA has initiated a string of Coordinated Research Programmes (CRPs) to enhance co-operation between fuel modellers. One of these CRPs, FUMEX-III, was ongoing during 2008 - 2011 and has provided material and incentive for assessment of the fuel codes. This report presents the Finnish FUMEX-III simulations performed with the VTT-modified ENIGMA v5.9b. The work has been done as a part of SAFIR2010 (SAfety of FInnish Reactors 2010) project POKEVA (years 2008 to 2010) and SAFIR2014 project PALAMA (year 2011). VTT Technical Research Centre of Finland received ENIGMA v5.9b from Nuclear Electric plc of the UK in 1992. Internal development has been on-going since then. 'ENIGMA v5.9b with VTT modifications' (from now on, 'ENIGMA' for short) is a separate and different program from British Energy's ENIGMA 5.14 and UK National Nuclear Laboratory's ENIGMA-B. The FUMEX-III work has been performed in tandem of VTT's internal review work attempting to catalogue the changes done since 1992 and to assess the current state of the code. Several individual internal reports detail the changes made and the individual model assessments done during the years. (author)

29 CFR Appendix III to Part 1918 - The Mechanics of Conventional Cargo Gear (Non-mandatory)

Science.gov (United States)

2010-07-01

... 29 Labor 7 2010-07-01 2010-07-01 false The Mechanics of Conventional Cargo Gear (Non-mandatory.... 1918, App. III Appendix III to Part 1918—The Mechanics of Conventional Cargo Gear (Non-mandatory) Note: This appendix is non-mandatory and provides an explanation of the mechanics in the correct spotting of...

Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

International Nuclear Information System (INIS)

Barranco R, F.

2015-01-01

In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

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

International Working Group on Past Reactors Thirteenth Annual Meeting. Summary Report. Part III

International Nuclear Information System (INIS)

1981-04-01

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

Experience with advanced driver fuels in EBR-II

International Nuclear Information System (INIS)

Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

1992-01-01

The Experimental Breeder Reactor II (EBR-II) is a complete nuclear power plant, incorporating a pool-type liquid-metal reactor (LMR) with a fuel-power thermal output of 62.5 MW and an electrical output of 20 MW. Initial criticality was in 1961, utilizing a metallic driver fuel design called the Mark-I. The fuel design has evolved over the last 30 yr, and significant progress has been made on improving performance. The first major innovations were incorporated into the Mark-II design, and burnup then increased dramatically. This design performed successfully, and fuel element lifetime was limited by subassembly hardware performance rather than the fuel element itself. Transient performance of the fuel was also acceptable and demonstrated the ability of EBR-II to survive severe upsets such as a loss of flow without scram. In the mid 1980s, with renewed interest in metallic fuels and Argonne's integral fast reactor (IFR) concept, the Mark-II design was used as the basis for new designs, the Mark-III and Mark-IV. In 1987, the Mark-III design began qualification testing to become a driver fuel for EBR-II. This was followed in 1989 by the Mark-IIIA and Mark-IV designs. The next fuel design, the Mark-V, is being planned to demonstrate the utilization of recycled fuel. The fuel cycle facility attached to EBR-II is being refurbished to produce pyroprocessed recycled fuel as part of the demonstration of the IFR

Towards an interpretation of the mechanism of the actinides(III)/lanthanides(III) separation by synergistic solvent extraction with nitrogen-containing polydendate ligands; Vers une interpretation des mecanismes de la separation actinides(III)/lanthanides(III) par extraction liquide-liquide synergique impliquant des ligands polyazotes

Energy Technology Data Exchange (ETDEWEB)

Francois, N [CEA/VALRHO - site de Marcoule, Dept. de Recherche en Retraitement et en Vitrification, (DRRV), 30 - Marcoule (France); Universite Henri Poincare, 54 - Vandoeuvre-les-Nancy (France)

2000-07-01

In the field of the separation of long-lived radionuclides from the wastes produced by nuclear fuel reprocessing, aromatic nitrogen-containing polydendate ligands are potential candidates for the selective extraction, alone or in synergistic mixture with acidic extractants, of trivalent actinides from trivalent lanthanides. The first part of this work deals with the complexation of trivalent f cations with various nitrogen-containing ligands (poly-pyridine analogues). Time-resolved laser-induced fluorimetry (TRLIF) and UV-visible spectrophotometry were used to determine the nature and evaluate the stability of each complex. Among the ligands studied, the least basic Me-Btp proved to be highly selective towards americium(III) in acidic solution. In the second part, two synergistic systems (nitrogen-containing polydendate ligand and lipophilic carboxylic acid) are studied and compared in regard to the extraction and separation of lanthanides(III) and actinides(III). TRLIF and gamma spectrometry allowed the nature of the extracted complexes and the optimal conditions of efficiency of both systems to be determined. Comparison between these different studies showed that the selectivity of complexation of trivalent f cations by a given nitrogen-containing polydendate ligand could not always be linked to the Am(III)Eu(III) selectivity reached in synergistic extraction. The latter depends on the 'balance' between the acid-basic properties on the one hand, and on the hard-soft characteristics on the other hand, of both components of synergistic system. (author)

Towards an interpretation of the mechanism of the actinides(III)/lanthanides(III) separation by synergistic solvent extraction with nitrogen-containing polydendate ligands; Vers une interpretation des mecanismes de la separation actinides(III)/lanthanides(III) par extraction liquide-liquide synergique impliquant des ligands polyazotes

Energy Technology Data Exchange (ETDEWEB)

Francois, N. [CEA/VALRHO - site de Marcoule, Dept. de Recherche en Retraitement et en Vitrification, (DRRV), 30 - Marcoule (France); Universite Henri Poincare, 54 - Vandoeuvre-les-Nancy (France)

2000-07-01

In the field of the separation of long-lived radionuclides from the wastes produced by nuclear fuel reprocessing, aromatic nitrogen-containing polydendate ligands are potential candidates for the selective extraction, alone or in synergistic mixture with acidic extractants, of trivalent actinides from trivalent lanthanides. The first part of this work deals with the complexation of trivalent f cations with various nitrogen-containing ligands (poly-pyridine analogues). Time-resolved laser-induced fluorimetry (TRLIF) and UV-visible spectrophotometry were used to determine the nature and evaluate the stability of each complex. Among the ligands studied, the least basic Me-Btp proved to be highly selective towards americium(III) in acidic solution. In the second part, two synergistic systems (nitrogen-containing polydendate ligand and lipophilic carboxylic acid) are studied and compared in regard to the extraction and separation of lanthanides(III) and actinides(III). TRLIF and gamma spectrometry allowed the nature of the extracted complexes and the optimal conditions of efficiency of both systems to be determined. Comparison between these different studies showed that the selectivity of complexation of trivalent f cations by a given nitrogen-containing polydendate ligand could not always be linked to the Am(III)Eu(III) selectivity reached in synergistic extraction. The latter depends on the 'balance' between the acid-basic properties on the one hand, and on the hard-soft characteristics on the other hand, of both components of synergistic system. (author)

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

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…

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...

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

ATWS: a reappraisal. Part III. Frequency of anticipated transients. Interim report

International Nuclear Information System (INIS)

Leverenz, F.L. Jr.; Koren, J.M.; Erdmann, R.C.; Lellouche, G.S.

1978-07-01

The document is Part III of the Institute study of the ATWS question. The frequencies of the various events which have led to a reactor scram are documented from the nuclear power plant records. Some of these events, in the absence of scram, could lead to undesirable system response and are the ''transients of significance'' which comprise the anticipated transients of the ATWS question

Design and initial performance of the Sandia Pulsed Reactor-III

International Nuclear Information System (INIS)

Reuscher, J.A.; Estes, B.F.

1976-01-01

The Sandia Pulsed Reactor-III (SPR-III) is a new fast pulsed reactor which has recently undergone initial testing at Sandia Laboratories. SPR-III is a uranium-10 weight percent molybdenum fuel assembly with a 17.78 cm irradiation cavity similar in design to SPR-II which has been in operation since 1967. The basic SPR-III design utilizes the same split-core configuration which has been proven with SPR-II; however, SPR-III uses external reflectors for control and external bolts to hold the fuel plates together. The core consists of sixteen fuel plates with an inside diameter of 17.78 cm, an outside diameter of 29.72 cm, and a core height of 31.9 cm. The fuel mass is about 227 kg of fully enriched uranium-10 weight percent molybdenum alloy. SPR III has completed the initial series of startup tests which included the critical experiment, zero and low-power tests, and pulse testing. The reactor design and results from the initial testing program are described in this paper. A portion of the startup experiments with SPR-III have been completed and this paper discusses the more important aspects of the initial testing program

The fuel cycle scoping system

International Nuclear Information System (INIS)

Dooley, G.D.; Malone, J.P.

1986-01-01

The Fuel Cycle Scoping System (FCSS) was created to fill the need for a scoping tool which provides the utilities with the ability to quickly evaluate alternative fuel management strategies, tails assay choices, fuel fabrication quotes, fuel financing alternatives, fuel cycle schedules, and other fuel cycle perturbations. The FCSS was specifically designed for PC's that support dBASE-III(TM), a relational data base software system by Ashton-Tate. However, knowledge of dBASE-III is not necessary in order to utilize the FCSS. The FCSS is menu driven and can be utilized as a teaching tool as well as a scoping tool

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

Uranium (III)-Plutonium (III) co-precipitation in molten chloride

Science.gov (United States)

Vigier, Jean-François; Laplace, Annabelle; Renard, Catherine; Miguirditchian, Manuel; Abraham, Francis

2018-02-01

Co-management of the actinides in an integrated closed fuel cycle by a pyrochemical process is studied at the laboratory scale in France in the CEA-ATALANTE facility. In this context the co-precipitation of U(III) and Pu(III) by wet argon sparging in LiCl-CaCl2 (30-70 mol%) molten salt at 705 °C is studied. Pu(III) is prepared in situ in the molten salt by carbochlorination of PuO2 and U(III) is then introduced as UCl3 after chlorine purge by argon to avoid any oxidation of uranium up to U(VI) by Cl2. The oxide conversion yield through wet argon sparging is quantitative. However, the preferential oxidation of U(III) in comparison to Pu(III) is responsible for a successive conversion of the two actinides, giving a mixture of UO2 and PuO2 oxides. Surprisingly, the conversion of sole Pu(III) in the same conditions leads to a mixture of PuO2 and PuOCl, characteristic of a partial oxidation of Pu(III) to Pu(IV). This is in contrast with coconversion of U(III)-Pu(III) mixtures but in agreement with the conversion of Ce(III).

Towards an interpretation of the mechanism of the actinides(III)/lanthanides(III) separation by synergistic solvent extraction with nitrogen-containing polydendate ligands

International Nuclear Information System (INIS)

Francois, N.

2000-01-01

In the field of the separation of long-lived radionuclides from the wastes produced by nuclear fuel reprocessing, aromatic nitrogen-containing polydendate ligands are potential candidates for the selective extraction, alone or in synergistic mixture with acidic extractants, of trivalent actinides from trivalent lanthanides. The first part of this work deals with the complexation of trivalent f cations with various nitrogen-containing ligands (poly-pyridine analogues). Time-resolved laser-induced fluorimetry (TRLIF) and UV-visible spectrophotometry were used to determine the nature and evaluate the stability of each complex. Among the ligands studied, the least basic Me-Btp proved to be highly selective towards americium(III) in acidic solution. In the second part, two synergistic systems (nitrogen-containing polydendate ligand and lipophilic carboxylic acid) are studied and compared in regard to the extraction and separation of lanthanides(III) and actinides(III). TRLIF and gamma spectrometry allowed the nature of the extracted complexes and the optimal conditions of efficiency of both systems to be determined. Comparison between these different studies showed that the selectivity of complexation of trivalent f cations by a given nitrogen-containing polydendate ligand could not always be linked to the Am(III)Eu(III) selectivity reached in synergistic extraction. The latter depends on the 'balance' between the acid-basic properties on the one hand, and on the hard-soft characteristics on the other hand, of both components of synergistic system. (author)

Neuroscience in Nazi Europe Part III: victims of the Third Reich.

Science.gov (United States)

Zeidman, Lawrence A; Kondziella, Daniel

2012-11-01

In Part I, neuroscience collaborators with the Nazis were discussed, and in Part II, neuroscience resistors were discussed. In Part III, we discuss the tragedy regarding european neuroscientists who became victims of the Nazi onslaught on “non-Aryan” doctors. Some of these unfortunate neuroscientists survived Nazi concentration camps, but most were murdered. We discuss the circumstances and environment which stripped these neuroscientists of their profession, then of their personal rights and freedom, and then of their lives. We include a background analysis of anti-Semitism and Nazism in their various countries, then discuss in depth seven exemplary neuroscientist Holocaust victims; including Germans Ludwig Pick, Arthur Simons, and Raphael Weichbrodt, Austrians Alexander Spitzer and Viktor Frankl, and Poles Lucja Frey and Wladyslaw Sterling. by recognizing and remembering these victims of neuroscience, we pay homage and do not allow humanity to forget, lest this dark period in history ever repeat itself.

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...

FEMAXI-III. An axisymmetric finite element computer code for the analysis of fuel rod performance

International Nuclear Information System (INIS)

Ichikawa, M.; Nakajima, T.; Okubo, T.; Iwano, Y.; Ito, K.; Kashima, K.; Saito, H.

1980-01-01

For the analysis of local deformation of fuel rods, which is closely related to PCI failure in LWR, FEMAXI-III has been developed as an improved version based on the essential models of FEMAXI-II, MIPAC, and FEAST codes. The major features of FEMAXI-III are as follows: Elasto-plasticity, creep, pellet cracking, relocation, densification, hot pressing, swelling, fission gas release, and their interrelated effects are considered. Contact conditions between pellet and cladding are exactly treated, where sliding or sticking is defined by iterations. Special emphasis is placed on creep and pellet cracking. In the former, an implicit algorithm is applied to improve numerical stability. In the latter, the pellet is assumed to be non-tension material. The recovery of pellet stiffness under compression is related to initial relocation. Quadratic isoparametric elements are used. The skyline method is applied to solve linear stiffness equation to reduce required core memories. The basic performance of the code has been proven to be satisfactory. (author)

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

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)

10 CFR Appendix III to Part 960 - Application of the System and Technical Guidelines During the Siting Process

Science.gov (United States)

2010-01-01

... 960—Application of the System and Technical Guidelines During the Siting Process 1. This appendix... 10 Energy 4 2010-01-01 2010-01-01 false Application of the System and Technical Guidelines During the Siting Process III Appendix III to Part 960 Energy DEPARTMENT OF ENERGY GENERAL GUIDELINES FOR THE...

Activation calculation of steel of the control rods of TRIGA Mark III reactor; Calculo de activacion del acero de las barras de control del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III) 2009–2014

International Nuclear Information System (INIS)

2015-10-01

At the beginning of 2014, there were 437 nuclear power reactors in operation and 72 reactors under construction. To date, around 370 500 t (HM) (tonnes of heavy metal) of spent fuel have been discharged from reactors, and approximately 253 700 t (HM) are stored at various storage facilities. Although wet storage at reactor sites still dominates, the amount of spent fuel being transferred to dry storage technologies has increased significantly since 2005. For example, around 28% of the total fuel inventory in the United States of America is now in dry storage. Although the licensing for the construction of geological disposal facilities is under way in Finland, France and Sweden, the first facility is not expected to be available until 2025 and for most States with major nuclear programmes not for several decades afterwards. Spent fuel is currently accumulating at around 7000 t (HM) per year worldwide. The net result is that the duration of spent fuel storage has increased beyond what was originally foreseen. In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. To address this, the IAEA continued the Coordinated Research Project (CRP) on Spent Fuel Performance Assessment and Research (SPAR-III) in 2009 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). This has been achieved through: evaluating surveillance and monitoring programmes of spent fuel and storage facilities; collecting and exchanging relevant experience of spent fuel storage and the impact on associated spent fuel management activities; facilitating the transfer of knowledge by documenting the technical basis for spent fuel storage; creating synergy among research projects of the participating Member States; and developing the capability to assess the impact

Further analysis of extended storage of spent fuel. Final report of a co-ordinated research programme on the behaviour of spent fuel assemblies during extended storage (BEFAST-III) 1991-1996

International Nuclear Information System (INIS)

1997-05-01

Considerable quantities of spent fuel continue to be produced and to accumulate in a number of countries. Although some new reprocessing facilities have been constructed, many countries are investigating the option of extended spent fuel storage prior to reprocessing or fuel disposal. Wet storage continues to predominate as an established technology. However, dry storage is becoming increasingly used with many countries considering dry storage for the longer term. This Technical Document is the final report of the IAEA Co-ordinated Research Programme on the Behaviour of Spent Fuel Assemblies During Extended Storage (BEFAST-III, 1991-1996). It contains analyses of wet and dry spent fuel storage technologies obtained from 16 organizations representing 13 countries (Canada, Finland, France, Germany, Hungary, the Republic of Korea, Japan, the Russian Federation, Slovakia, Spain, Sweden, the United Kingdom and the USA) which participated in the co-ordinated research programme as participants or observers. The report contains information presented during the three Research Co-ordination meetings and also data which were submitted by the participants in response to request by the Scientific Secretary. 48 refs, 4 tabs

Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

International Nuclear Information System (INIS)

Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

2003-01-01

The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

These affect the mechanical design of the fuel assembly components. And thus, appropriate structural design criteria should also be chosen to incorporate the specific design conditions of the SFR fuel assemblies. Among them, the temperature is one of the most crucial conditions to be concerned because the sodium coolant temperature is normally more than 500ºC which is much higher than that of the LWR (< 350ºC). This implies that a thermal creep should be significantly considered in the SFR fuel assembly mechanical design. In addition to the high temperature condition, an irradiation swelling is also an important behavior that the SFR fuel assembly material should accommodate. To incorporate the temperature and irradiation impacts, the material of the fuel assembly components is presently determined to be made of HT-9, the ferriticmartensitic steel. In this paper, the ASME Sec. III Div. 5 (referred to as ‘Div. 5’ hereinafter), which was developed for a ‘high temperature reactor’, is considered as one of the structural design criteria for the mechanical design of SFR fuel assemblies. In this paper, the stress intensity limits, S{sub m} and S{sub t} of HT-9 were built for the structural criteria of an SFR fuel assembly. S{sub m} is obtained from the ultimate strength. As for S{sub t}, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of S{sub mt}, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as S{sub mt} under the temperature about 470ºC which is relatively low temperature range and over 470ºC with relatively short time duration as 1000 hours. And the S{sub t} is adopted as Smt at over 470ºC and long time duration over 34800 hours, and over 520ºC and 10{sup 4} hours too. And at over 570ºC and 1000 hours, and at over 630ºC and 100 hours, S{sub t} is also adopted for S{sub mt}.

The evaluation of failure stress and released amount of fission product gas of power ramped rod by fuel behaviour analysis code 'FEMAXI-III'

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujita, Misao

1984-01-01

Pellet-Cladding Interaction(PCI) related in-pile failure of Zircaloy sheathed fuel rod is in general considered to be caused by combination of pellet-cladding mechanical interaction(PCMI) with fuel-cladding chemical interaction(FCCI). An understanding of a basic mechanism of PCI-related fuel failure is therefore necessary to get actual cladding hoop stress from mechanical interaction and released amounts of fission product(FP) gas of aggressive environmental agency from chemical interaction. This paper describes results of code analysis performed on fuel failure to cladding hoop stress and amounts of FP gas released under the condition associated with power ramping. Data from Halden(HBWR) and from Studsvik(R2) are used for code analysis. The fuel behaviour analysis code ''FEMAXI-III'' is used as an analytical tool. The followings are revealed from the study: (1) PCI-related fuel failure is dependent upon cladding hoop stress and released amounts of FP gas at power ramping. (2) Preliminary calculated threshold values of hoop stress and of released amounts of FP gas to PCI failure are respectively 330MPa, 10% under the Halden condition, 190MPa, 5% under the Inter ramp(BWR) condition, and 270MPa, 14% under the Over ramp(PWR) condition. The values of hoop stress calculated are almost in the similar range of those obtained from ex-reactor PCI simulated tests searched from references published. (3) The FEMAXI-III code verification is made in mechanical manner by using in-pile deformation data(diametral strain) obtained from power ramping test undertaken by JAERI. While, the code verification is made in thermal manner by using punctured FP gas data obtained from post irradiation examination performed on non-defected power ramped fuel rods. The calculations are resulted in good agreements to both, mechanical and thermal experimental data suggesting the validity of the code evaluation. (J.P.N.)

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

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

Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

International Nuclear Information System (INIS)

Tome, Carlos N.; Caro, J.A.; Lebensohn, R.A.; Unal, Cetin; Arsenlis, A.; Marian, J.; Pasamehmetoglu, K.

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.

Insight into the Extraction Mechanism of Americium(III) over Europium(III) with Pyridylpyrazole: A Relativistic Quantum Chemistry Study.

Science.gov (United States)

Kong, Xiang-He; Wu, Qun-Yan; Wang, Cong-Zhi; Lan, Jian-Hui; Chai, Zhi-Fang; Nie, Chang-Ming; Shi, Wei-Qun

2018-05-10

Separation of trivalent actinides (An(III)) and lanthanides (Ln(III)) is one of the most important steps in spent nuclear fuel reprocessing. However, it is very difficult and challenging to separate them due to their similar chemical properties. Recently the pyridylpyrazole ligand (PypzH) has been identified to show good separation ability toward Am(III) over Eu(III). In this work, to explore the Am(III)/Eu(III) separation mechanism of PypzH at the molecular level, the geometrical structures, bonding nature, and thermodynamic behaviors of the Am(III) and Eu(III) complexes with PypzH ligands modified by alkyl chains (Cn-PypzH, n = 2, 4, 8) have been systematically investigated using scalar relativistic density functional theory (DFT). According to the NBO (natural bonding orbital) and QTAIM (quantum theory of atoms in molecules) analyses, the M-N bonds exhibit a certain degree of covalent character, and more covalency appears in Am-N bonds compared to Eu-N bonds. Thermodynamic analyses suggest that the 1:1 extraction reaction, [M(NO 3 )(H 2 O) 6 ] 2+ + PypzH + 2NO 3 - → M(PypzH)(NO 3 ) 3 (H 2 O) + 5H 2 O, is the most suitable for Am(III)/Eu(III) separation. Furthermore, the extraction ability and the Am(III)/Eu(III) selectivity of the ligand PypzH is indeed enhanced by adding alkyl-substituted chains in agreement with experimental observations. Besides this, the nitrogen atom of pyrazole ring plays a more significant role in the extraction reactions related to Am(III)/Eu(III) separation compared to that of pyridine ring. This work could identify the mechanism of the Am(III)/Eu(III) selectivity of the ligand PypzH and provide valuable theoretical information for achieving an efficient Am(III)/Eu(III) separation process for spent nuclear fuel reprocessing.

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

Social class, political power, and the state: their implications in medicine--part III.

Science.gov (United States)

Navarro, V

1977-01-01

This is the third part of an article on the distribution of power and the nature of the state in Western industrialized societies and their implications in medicine. Parts I and II were published in the preceding issue of this Journal. Part I presented a critique of contemporary theories of the Western system of power; discussed the countervailing pluralist and power of elite theories, as well as those of bureaucratic and professional control; and concluded with an examination of the Marxist theories of economic determinism, structural determinism, and corporate statism. Part II presented a Marxist theory of the role, nature, and characteristics of state intervention. Part III focuses on the mode of that intervention and the reasons for its growth, with an added analysis of the attributes of state intervention in the health sector, and of the dialectical relationship between its growth and the current fiscal crisis of the state. In all three parts, the focus is on Western European countries and on North America, with many examples and categories from the area of medicine.

Dry process fuel performance technology development

International Nuclear Information System (INIS)

Kang, Kweon Ho; Kim, K. W.; Kim, B. K.

2006-06-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the 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 III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, 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 (KAOS)

Dry process fuel performance technology development

Energy Technology Data Exchange (ETDEWEB)

Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

2006-06-15

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the 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 III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, 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 (KAOS)

Gen-III/III+ reactors. Solving the future energy supply shortfall. The SWR-1000 option

International Nuclear Information System (INIS)

Stosic, Z.V.

2006-01-01

Deficiency of non-renewable energy sources, growing demand for electricity and primary energy, increase in population, raised concentration of greenhouse gases in the atmosphere and global warming are the facts which make nuclear energy currently the most realistic option to replace fossil fuels and satisfy global demand. The nuclear power industry has been developing and improving reactor technology for almost five decades and is now ready for the next generation of reactors which should solve the future energy supply shortfall. The advanced Gen-III/III+ (Generation III and/or III+) reactor designs incorporate passive or inherent safety features which require no active controls or operational intervention to manage accidents in the event of system malfunction. The passive safety equipment functions according to basic laws of physics such as gravity and natural convection and is automatically initiated. By combining these passive systems with proven active safety systems, the advanced reactors can be considered to be amongst the safest equipment ever made. Since the beginning of the 90's AREVA NP has been intensively engaged in the design of two advanced Gen-III+ reactors: (i) PWR (Pressurized Water Reactor) EPR (Evolutionary Power Reactor) and (ii) BWR (Boiling Water Reactor) SWR-1000. The SWR-1000 reactor design marks a new era in the successful tradition of BWR technology. It meets the highest safety standards, including control of a core melt accident. This is achieved by supplementing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation. A short construction period, flexible fuel cycle lengths and a high fuel discharge burn-up contribute towards meeting economic goals. The SWR-1000 completely fulfils international nuclear regulatory requirements. (author)

Development of failed fuel detection system for PWR (III)

International Nuclear Information System (INIS)

Hwang, Churl Kew; Kang, Hee Dong; Jeong, Seung Ho; Cho, Byung Sub; Yoon, Byeong Joo; Yoon, Jae Seong

1987-12-01

Ultrasonic transducers satisfying the conditions for failed fuel rod detection for failed fuel rod detection have been designed and built. And performance tests for them have been carried out. Ultrasonic signal processing units, a manipulator guiding the ultrasonic probe through the fuel assembly lanes and its control units have been constructed. The performance of the system has been verified experimentally to be successful in failed fuel rod detection. (Author)

Thermal sensation and comfort models for non-uniform and transient environments: Part III: whole-body sensation and comfort

OpenAIRE

Zhang, Hui; Arens, Edward; Huizenga, Charlie; Han, Taeyoung

2009-01-01

A three-part series presents the development of models for predicting the local thermal sensation (Part I) and local thermal comfort (Part II) of different parts of the human body, and also the whole-body sensation and comfort (Part III) that result from combinations of local sensation and comfort. The models apply to sedentary activities in a range of environments: uniform and non-uniform, stable and transient. They are based on diverse findings from the literature and from body-part-specifi...

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

Activation calculation of steel of the control rods of TRIGA Mark III reactor

International Nuclear Information System (INIS)

Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A.

2014-10-01

In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

Uranium (III) precipitation in molten chloride by wet argon sparging

Energy Technology Data Exchange (ETDEWEB)

Vigier, Jean-François, E-mail: jean-francois.vigier@ec.europa.eu [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France); Laplace, Annabelle [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Renard, Catherine [Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France); Miguirditchian, Manuel [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Abraham, Francis [Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France)

2016-06-15

In the context of pyrochemical processes for nuclear fuel treatment, the precipitation of uranium (III) in molten salt LiCl-CaCl{sub 2} (30–70 mol%) at 705 °C is studied. First, this molten chloride is characterized with the determination of the water dissociation constant. With a value of 10{sup −4.0}, the salt has oxoacid properties. Then, the uranium (III) precipitation using wet argon sparging is studied. The salt is prepared using UCl{sub 3} precursor. At the end of the precipitation, the salt is totally free of solubilized uranium. The main part is converted into UO{sub 2} powder but some uranium is lost during the process due to the volatility of uranium chloride. The main impurity of the resulting powder is calcium. The consequences of oxidative and reductive conditions on precipitation are studied. Finally, coprecipitation of uranium (III) and neodymium (III) is studied, showing a higher sensitivity of uranium (III) than neodymium (III) to precipitation. - Highlights: • Precipitation of Uranium (III) is quantitative in molten salt LiCl-CaCl{sub 2} (30–70 mol%). • The salt is oxoacid with a water dissociation constant of 10{sup −4.0} at 705 °C. • Volatility of uranium chloride is strongly reduced in reductive conditions. • Coprecipitation of U(III) and Nd(III) leads to a consecutive precipitation of the two elements.

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.

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-11-15

This volume includes the following reports: radiation damage of metal uranium; influence of burnup rate on the stability of metal uranium fuel, influence of precipitation and desorption of inert gas on the density change of samples.

Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

Energy Technology Data Exchange (ETDEWEB)

Barranco R, F.

2015-07-01

In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

Application of fuel management calculation codes for CANDU reactor

International Nuclear Information System (INIS)

Ju Haitao; Wu Hongchun

2003-01-01

Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

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

Theoretical analysis of nuclear reactors (Phase III), I-V, Part IV, Influence of isotopic composition of nuclear fuel on the reactivity with constant flux; Razrada metoda teorijske analize nuklearnih reaktora (III faza) I-IV, IV Deo, Uticaj promene izotopnog sastava goriva na reaktivnost uz konstantan fluks

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-01-15

Part one of this report presents a series of differential equations describing the nuclear fuel depletion during reactor operation. This series of differential equations is extended to describe the fission products. This part includes equations for effective multiplication factor k{sub eff} and reactivity {rho} as a function of irradiation {tau}. Part two includes results obtained on the analog computer PACE 231 R, and related to Calder Hall type reactor. Part three covers detailed preparation of the series of equations for solution by using the analog computer. Part four includes the list of references related to this task.

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

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

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

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

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...

Final Report for the FUMEX-III Exercise with the TRANSURANUS Code

International Nuclear Information System (INIS)

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

2013-01-01

This report describes the results of the fourth round robin exercise organized by the IAEA for the LWR fuel behaviour codes. The previous exercise was organized by the IAEA in 2002-2007. ITU contributed to the IAEA FUMEX- II co-ordinated research project (CRP) by: - The development of a high burn-up TRANSURANUS-WWER Version - Verification of the TRANSURANUS-WWER Version for WWER-1000 reactors - Further verification of the TRANSURANUS Code by selected irradiations from the IFPE Database - Transfer of latest ITU knowledge in the following areas: high burnup effects and MOX behaviour as far as confidentiality is not concerned. Within the FUMEX-III project this work continues by: - Knowledge transfer and release of the TRANSURANUS code to safety authorities in several neighbouring countries of the European Union - Further verification of the TRANSURANUS Code by selected irradiations from the IFPE Database: extending the verification of the TRANSURANUS code on the basis of the cases related to the behaviour of high burnup UO 2 fuel, Gd-containing UO 2 fuel and MOX fuel under normal operating conditions in LWRs, including WWER. The simulation of the pellet-cladding mechanical interaction received particular attention. - Extending the TRANSURANUS code for simulation of fuel rods under accidental conditions, such as a loss of coolant accidents (LOCA) and a Reactivity Initiated Accident (RIA), for which various code improvements have been implemented and other changes are still under development. During the entire period of the CRP, ITU has carried out all priority cases, except for IFA-519 for which insufficient data have been provided. All these results are considered in this report. However, in order to limit the redundancy of work of various TRANSURANUS users involved in the FUMEX-III CRP, ITU has co-ordinated part of the work of the partners in Bulgaria, Romania and Italy. More precisely, the work on the WWER version of the code has been carried out in collaboration

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…

Improvement of Computer Codes Used for Fuel Behaviour Simulation (FUMEX-III). Report of a Coordinated Research Project 2008-2012

International Nuclear Information System (INIS)

2013-03-01

It is fundamental to the future of nuclear power that reactors can be run safely and economically to compete with other forms of power generation. As a consequence, it is essential to develop the understanding of fuel performance and to embody that knowledge in codes to provide best estimate predictions of fuel behaviour. This in turn leads to a better understanding of fuel performance, a reduction in operating margins, flexibility in fuel management and improved operating economics. The IAEA has therefore embarked on a series of programmes addressing different aspects of fuel behaviour modelling with the following objectives: - To assess the maturity and prediction capabilities of fuel performance codes, and to support interaction and information exchange between countries with code development and application needs (FUMEX series); - To build a database of well defined experiments suitable for code validation in association with the OECD Nuclear Energy Agency (OECD/NEA); - To transfer a mature fuel modelling code to developing countries, to support teams in these countries in their efforts to adapt the code to the requirements of particular reactors, and to provide guidance on applying the code to reactor operation and safety assessments; - To provide guidelines for code quality assurance, code licensing and code application to fuel licensing. This report describes the results of the coordinated research project on the ''Improvement of computer codes used for fuel behaviour simulation (FUMEX-III)''. This programme was initiated in 2008 and completed in 2012. It followed previous programmes on fuel modelling: D-COM 1982-1984, FUMEX 1993-1996 and FUMEX-II 2002-2006. The participants used a mixture of data derived from commercial and experimental irradiation histories, in particular data designed to investigate the mechanical interactions occurring in fuel during normal, transient and severe transient operation. All participants carried out calculations on priority

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)

Fracture toughness of A533B Part III - variability of A533B fracture toughness as determined from Charpy data

International Nuclear Information System (INIS)

Druce, S.G.; Eyre, B.L.

1978-08-01

This is the final part of a series of three reports examining the upper shelf fracture toughness of A533B Class 1 pressure vessel steel. Part I (AERE R 8968) critically reviews the current elasto plastic fracture mechanics methodologies employed to characterise toughness following extensive yielding and Part II (AERE R 8969) examines several sources of fracture mechanics data pertinent to A533B Class 1 in the longitudinal (RW) orientation. Part III is a review of the effects of (i) position and orientation within the plate (ii) welding processes and post weld heat treatment and (iii) neutron irradiation as measured by Charpy impact testing. It is concluded that the upper shelf factor energy is dependent on orientation and position and can be reduced by welding, extended post weld heat treatments and neutron irradiation. Neutron irradiation effects are known to be strongly dependent on composition and metallurgical conditions, but an explanation for the variability following extended post weld treatments has yet to be resolved. (author)

Active Control of Low-Speed Fan Tonal Noise Using Actuators Mounted in Stator Vanes: Part III Results

Science.gov (United States)

Sutliff, Daniel L.; Remington, Paul J.; Walker, Bruce E.

2003-01-01

A test program to demonstrate simplification of Active Noise Control (ANC) systems relative to standard techniques was performed on the NASA Glenn Active Noise Control Fan from May through September 2001. The target mode was the m = 2 circumferential mode generated by the rotor-stator interaction at 2BPF. Seven radials (combined inlet and exhaust) were present at this condition. Several different error-sensing strategies were implemented. Integration of the error-sensors with passive treatment was investigated. These were: (i) an in-duct linear axial array, (ii) an induct steering array, (iii) a pylon-mounted array, and (iv) a near-field boom array. The effect of incorporating passive treatment was investigated as well as reducing the actuator count. These simplified systems were compared to a fully ANC specified system. Modal data acquired using the Rotating Rake are presented for a range of corrected fan rpm. Simplified control has been demonstrated to be possible but requires a well-known and dominant mode signature. The documented results here in are part III of a three-part series of reports with the same base title. Part I and II document the control system and error-sensing design and implementation.

Design studies for the Mark-III core of experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu; Shindo, Ryuiti; Arai, Taketoshi

1979-08-01

The Mark-III core in the first conceptual design made in 1975 is a fundamental core for VHTR. Subsequently, further design studies were made fuel loading scheme and control rod withdrawal sequence for the core to increase its safety margin (shutdown margin, etc.) and operational margin (minimum Reynolds number, maximum fuel temperature, etc.). It was shown that the Mark-III should exhibit the performance expected of VHTR, unless changes are made in the preconditions for its nuclear, thermal-hydraulic design. Also, the needs as below were indicated: (1) reasonable core design criteria and guidelines, (2) fuel-loading-scheme requirements in fuel management, fuel misloading and reactor operation, (3) confirmation on precision of the core design method and its further refinement. (author)

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…

Visual inspection system and sipping design for spent fuel at TRIGA MARK III reactor of Mexico

International Nuclear Information System (INIS)

Delfin, A.; Mazon, R.

2002-01-01

In the framework of the Technical Cooperation Regional Project for Latin America RLA/4/018 for the biennium 2001-2002, one of the activities identified is the characterization of spent fuel. Of these activities an important one is not doubt the physical condition of spent fuel because an appropriate identification of the fuel status will prevent problems of fuel leaks, corrosion problems etc. As part of the activities of the project was decided that countries no having visual inspection and sipping systems should be very desirable to have them as a result of this project. The Triga reactor of Mexico does not have both of them, therefore, it was decided the need of having both system. The paper describe first the way we designed and constructed a remote Visual Inspection System and example of how is operated. Along the experience and problems we have had with the system. Also we will present the design of the Sipping system were two option were considered. First to take a sample of water after a convenient period of time passing through a circuit to a multichannel analyzer and to identify leakage by way of measuring Caesium-137. Second, exists the possibility that the Stainless Steel sleeve of the fuel has only very small failures, so it is going to be very difficult to have leakages unless the fuel is hot. Therefore we are evaluating the possibility of using heaters to increase the temperature of the fuel and succeed on detecting leakages. The results - we hope - will be ready to be presented at the meeting. (author)

Thallium (III) salts utilization in organic synthesis. Part II

International Nuclear Information System (INIS)

Ferraz, H.M.C.

1989-01-01

The utilizations of thallium (III) salts in organic synthesis with carbonylic and acitylenic substrates are presented. The reactions of carbonylic substra3ts with kitones and the oxidation reactions of acetylenic substrates are shown. Others reactions including thallium (III) salts and non aromatic unsatured substracts, as cleasage of ethers and epoxide using thallium trinitrate, hydrazones treatments with thallium triacetates, etc, are also mentioned. (C.G.C.) [pt

Building human resources capability in health care: a global analysis of best practice--Part III.

Science.gov (United States)

Zairi, M

1998-01-01

This is the last part of a series of three papers which discussed very comprehensively best practice applications in human resource management by drawing special inferences to the healthcare context. It emerged from parts I and II that high performing organisations plan and intend to build sustainable capability through a systematic consideration of the human element as the key asset and through a continuous process of training, developing, empowering and engaging people in all aspects of organisational excellence. Part III brings this debate to a close by demonstrating what brings about organisational excellence and proposes a road map for effective human resource development and management, based on world class standards. Healthcare human resource professionals can now rise to the challenge and plan ahead for building organisational capability and sustainable performance.

Molten fuel motion during a fast-reactor overpower transient

International Nuclear Information System (INIS)

Kolesar, D.C.; Padilla, A. Jr.; Lewis, C.H.; Waltar, A.E.

1976-01-01

Mechanistic models for postfailure fuel behavior during hypothetical transient overpower accidents are currently being developed for incorporation into the MELT accident analysis code. A new model for the fuel-coolant interaction and for the motion of fuel in the coolant channel has been developed and incorporated into the MELT-III code. A major limitation of the mechanistic fuel motion model is its dependence on the uniform interaction region of MELT-III. Consequently, a parallel effort is currently in progress to incorporate a non-uniform interaction region into the MELT code. Combination of the fuel motion and the nonuniform interaction region models will provide the framework for development of a mechanistic fuel plateout/blockage model for transient overpower accidents

Study of the chemical behaviour of technetium during irradiated fuels reprocessing

International Nuclear Information System (INIS)

Zelverte, A.

1988-04-01

This paper deals with the preparation of the lower oxidation states +III +IV and +V of technetium in nitric acid and its behaviour during the reprocessing of nuclear fuels (PUREX process). The first part of this work is a bibliographical study of this element in solution without any strong ligand. By chemical and electrochemical technics, pentavalent, tetravalent and trivalent technetium species, were prepared in nitric acid. The following chemical reactions are studied: - trivalent and tetravalent technetium oxidation by nitrate ion. - hydrazine and tetravalent uranium oxidation catalysed by technetium: in those reactions, we point out unequivocally the prominent part of trivalent and tetravalent technetium, - technetium behaviour towards hydroxylamine. Technetium should not cause any disturbance in the steps where hydroxylamine is employed to destroy nitrous acid and hydrazine replacement by hydroxylamine in uranium-plutonium partition could contribute to a best reprocessing of nuclear fuels [fr

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

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

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.)

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.

Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

2012-10-01

Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

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.)

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

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.

Liquid-liquid extraction kinetics of uranyl nitrate and actinides (III)-lanthanides nitrates by extractants with amide function

International Nuclear Information System (INIS)

Toulemonde, V.

1995-01-01

Nowadays, the most important part of electric power is generated by fission energy. But spent fuels have then to be reprocessed. The production of these reprocessed materials separately and with a high purity level is done according to a liquid-liquid extraction process (Purex process) with the use of tributyl phosphate as solvent. Optimization studies concerning the extracting agent have been undertaken. This work gives the results obtained for the uranyl nitrate and the actinides (III)-lanthanides (III) nitrates extraction by extractants with amide function (monoamide for U(VI) and diamide for actinides (III) and lanthanides (III)). The extraction kinetics have been studied in the case of a metallic specie transfer from the aqueous phase towards the organic phase. The experiments show that the nitrates extraction kinetics is limited by the complexation chemical reaction of the species at the interface between the two liquids. An adsorption-desorption interfacial reactional mechanism (Langmuir theory) is proposed for the uranyl nitrate. (O.M.)

Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

International Nuclear Information System (INIS)

Hernandez G, J.

2012-10-01

Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

Japanese contributions to IAEA INTOR workshop, phase two A, part 2, chapter III: impurity control (engineering)

International Nuclear Information System (INIS)

Seki, Masahiro; Miki, Nobuharu; Shibutani, Yoji; Fujimura, Kaoru; Adachi, Jun-ichi; Sato, Kosuke; Fujii, Masaharu; Yamazaki, Seiichiro; Itoh, Shin-ichi.

1985-07-01

This report corresponds to the second half of Chapter III of Japanese contribution report to IAEA INTOR Workshop, Phase Two A, Part 2. Data base assessment are made on candidate materials for the divertor, limiter, and the first wall. Engineering trade-off studies are made for the high-recycling and low temperature conditions. The studies include material considerations, configuration, thermohydraulic and stress analysis, disruption, lifetime analysis, and tritium permeation. (author)

76 FR 5319 - Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline

Science.gov (United States)

2011-01-31

... Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline AGENCY: Environmental... gasoline. This proposed rule will provide flexibility to the regulated community by allowing an additional... A. Alternative Test Method for Olefins in Gasoline III. Statutory and Executive Order Reviews A...

76 FR 65382 - Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline

Science.gov (United States)

2011-10-21

... Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline AGENCY: Environmental... gasoline. This final rule will provide flexibility to the regulated community by allowing an additional... Method for Olefins in Gasoline III. Statutory and Executive Order Reviews A. Executive Order 12866...

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

1977-04-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

International Nuclear Information System (INIS)

Boltax, A.; Biancheria, A.

1977-01-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

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

Emission factors of air pollutants from CNG-gasoline bi-fuel vehicles: Part II. CO, HC and NOx.

Science.gov (United States)

Huang, Xiaoyan; Wang, Yang; Xing, Zhenyu; Du, Ke

2016-09-15

The estimation of emission factors (EFs) is the basis of accurate emission inventory. However, the EFs of air pollutants for motor vehicles vary under different operating conditions, which will cause uncertainty in developing emission inventory. Natural gas (NG), considered as a "cleaner" fuel than gasoline, is increasingly being used to reduce combustion emissions. However, information is scarce about how much emission reduction can be achieved by motor vehicles burning NG (NGVs) under real road driving conditions, which is necessary for evaluating the environmental benefits for NGVs. Here, online, in situ measurements of the emissions from nine bi-fuel vehicles were conducted under different operating conditions on the real road. A comparative study was performed for the EFs of black carbon (BC), carbon monoxide (CO), hydrocarbons (HCs) and nitrogen oxides (NOx) for each operating condition when the vehicles using gasoline and compressed NG (CNG) as fuel. BC EFs were reported in part I. The part II in this paper series reports the influence of operating conditions and fuel types on the EFs of CO, HC and NOx. Fuel-based EFs of CO showed good correlations with speed when burning CNG and gasoline. The correlation between fuel-based HC EFs and speed was relatively weak whether burning CNG or gasoline. The fuel-based NOx EFs moderately correlated with speed when burning CNG, but weakly correlated with gasoline. As for HC, the mileage-based EFs of gasoline vehicles are 2.39-12.59 times higher than those of CNG vehicles. The mileage-based NOx EFs of CNG vehicles are slightly higher than those of gasoline vehicles. These results would facilitate a detailed analysis of the environmental benefits for replacing gasoline with CNG in light duty vehicles. Copyright © 2016 Elsevier B.V. All rights reserved.

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.)

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.

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.

Reactions of sigma-bonded organochromium(III)complexes

International Nuclear Information System (INIS)

Leslie, J.P. II.

1975-12-01

Three projects were carried out, each dealing with the kinetics and mechanism of reactions of sigma-bonded organochromium(III) complexes of the form (H 2 O) 5 CrR 2+ . Part I describes the kinetics of the reaction of dichloromethylchromium(III) ion with chromium(II) ion in aqueous acid. Part II deals with the radioexchange of 4-pyridinomethylchromium(III) ion with 51 Cr 2+ and the kinetics of formation of the organochromium species at 55 0 in 1 M H + . Part III deals with the reactions of Hg 2+ and CH 3 Hg + with a series of (H 2 O) 5 CrR 2+ complexes, in which R is an aliphatic alkyl group, a haloalkyl group, or an aralkyl group

Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

International Nuclear Information System (INIS)

Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

2016-01-01

Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

Environmental, health, and safety issues of fuel cells in transportation. Volume 1: Phosphoric acid fuel-cell buses

Energy Technology Data Exchange (ETDEWEB)

Ring, S

1994-12-01

The U.S. Department of Energy (DOE) chartered the Phosphoric Acid Fuel-Cell (PAFC) Bus Program to demonstrate the feasibility of fuel cells in heavy-duty transportation systems. As part of this program, PAFC- powered buses are being built to meet transit industry design and performance standards. Test-bed bus-1 (TBB-1) was designed in 1993 and integrated in March 1994. TBB-2 and TBB-3 are under construction and should be integrated in early 1995. In 1987 Phase I of the program began with the development and testing of two conceptual system designs- liquid- and air-cooled systems. The liquid-cooled PAFC system was chosen to continue, through a competitive award, into Phase H, beginning in 1991. Three hybrid buses, which combine fuel-cell and battery technologies, were designed during Phase III. After completing Phase II, DOE plans a comprehensive performance testing program (Phase HI) to verify that the buses meet stringent transit industry requirements. The Phase III study will evaluate the PAFC bus and compare it to a conventional diesel bus. This NREL study assesses the environmental, health, and safety (EH&S) issues that may affect the commercialization of the PAFC bus. Because safety is a critical factor for consumer acceptance of new transportation-based technologies the study focuses on these issues. The study examines health and safety together because they are integrally related. In addition, this report briefly discusses two environmental issues that are of concern to the Environmental Protection Agency (EPA). The first issue involves a surge battery used by the PAFC bus that contains hazardous constituents. The second issue concerns the regulated air emissions produced during operation of the PAFC bus.

A method for the preparation of a fuel, by the addition of one or more components to a base fuel

NARCIS (Netherlands)

2013-01-01

The present invention relates to a method for the preparation of a fuel, by the addition of one or more components to a base fuel, wherein the method comprises the following steps: i) providing a base fuel; ii) withdrawing aromatic components from a styrene / propylene ox ide production plant; iii)

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.)

Tanzania 1895-1920 : Part III: 1914-1920s

NARCIS (Netherlands)

Dietz, A.J.

2016-01-01

An earlier version of this African Postal Heritage Paper was published as African Studies Centre Leiden Working Paper 119 / 2015: "A postal history of the First World War in Africa and its aftermath - German colonies; III Deutsch Ostafrika / German East Africa", written by Ton Dietz.

Extraction of Am (III) and Nd (III): comparison of TODGA and TEHDGA

International Nuclear Information System (INIS)

Gujar, R.B.; Murali, M.S.; Ansari, S.A.; Manchanda, V.K.

2009-01-01

Belonging to the class of extractants, diglycolamides which are recently explored and promising for actinide partitioning, two reagents (N, N, N', N'-tetraoctyl diglycolamide) TODGA and its isomerically substituted counterpart, (N, N, N', N'- tetraethylhexyl diglycolamide) TEHDGA after addition of suitable phase modifiers, Dihexyoctanamide and isodecanol respectively in dodecane have been compared in their extraction abilities for Am (III) and Nd (III) from nitric acid as well as simulated high-level waste solutions (SHLW) equivalent to HLW arising from PHWR fuel reprocessing. Both 0.1M TODGA + 0.5M DHOA and 0.2M TEHDGA + 30% isodecanol in dodecane display high distribution ratios for the trivalent metal ions of f-elements. Similarities and differences in their extraction are discussed. (author)

Nonlinear observer designs for fuel cell power systems

Science.gov (United States)

Gorgun, Haluk

dynamics, and estimate not only hydrogen but also all other species in its reactors. We design nonlinear observers for the Catalytic Partial Oxidation (CPO), Water Gas Shift (WGS), and Preferential Oxidation (PROX), reactors in the FPS. The observers make use of temperature measurements (and possibly one more variable, such as pressure) to estimate the mole fractions of each species in the reactors. An advantage of these designs is that they are based on reaction invariants and do not rely on knowledge of reaction rate expressions. Finally, in part III, we illustrate how the designs of parts I and II can be incorporated in fault detection and estimation algorithms for common failures encountered in fuel cells, such as the cathode blower failure and the anode valve failure. For this task, we combine geometric tools with our observers.

Separation by liquid-liquid extraction of actinides(III) from lanthanides(III) using new molecules: the picolinamides; Separation par extraction liquide-liquide des actinides(III) des lanthanides(III) par de nouvelles molecules: les picolinamides

Energy Technology Data Exchange (ETDEWEB)

Cordier, P Y [CEA Marcoule, Departement de Recherche en Retraitement et en Vitrification, 30 - Bagnols-sur-Ceze (France); [Clermont-Ferrand-2 Univ., 63 - Aubiere (France)

1996-07-01

In the field of long-lived radionuclides separation from waste generated during spent fuel reprocessing, the picolinamides have been chosen as potential extractants for the selective extraction of actinides (III) from lanthanides (III). The first studies initiated on the most simple molecule of the picolinamide family, namely 2-pyridinecarboxamide, pointed out that in an aqueous media the complexation stability constant between this ligand and Am(III) is roughly 10 times higher than the ones corresponding to Ln(III). The synthesis of lipophilic derivatives of 2-pyridinecarboxamide leaded to extraction experiments. The extraction of metallic cation by lipophilic picolinamides, according to a solvatation mechanism, is strongly dependent on the nature of the amide function: a primary amide function (group I) leads to a good extraction; on the contrary, there is a decrease for secondary (group II) and tertiary (group III) amide functions. From a theoretical point of view, this work leads finally to the following conclusions: confirmation of the importance of the presence of soft donor atoms within the extractants (nitrogen in our case) for An(III)/Ln(III). Also, sensitivity of this soft donor atom regarding the protonation reaction; prevalence in our case of the affinity of the extractant for the metallic cation over the lipophilia of the extractant to ensure good distribution coefficients. The extraction and Am(III)/Ln(III) separation performances of the picolinamides from pertechnetic media leads to the design of a possible flowsheet for the reprocessing of high level liquid waste, with the new idea of an integrated technetium reflux. (author) 105 refs.

Issues of high-burnup fuel for advanced nuclear reactors

International Nuclear Information System (INIS)

Belac, J.; Milisdoerfer, L.

2004-12-01

A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

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

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).

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.

Impact of Fe(III) as an effective electron-shuttle mediator for enhanced Cr(VI) reduction in microbial fuel cells: Reduction of diffusional resistances and cathode overpotentials

Energy Technology Data Exchange (ETDEWEB)

Wang, Qiang [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Huang, Liping, E-mail: lipinghuang@dlut.edu.cn [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Pan, Yuzhen [College of Chemistry, Dalian University of Technology, Dalian 116024 (China); Quan, Xie [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Li Puma, Gianluca, E-mail: g.lipuma@lboro.ac.uk [Environmental Nanocatalysis & Photoreaction Engineering, Department of Chemical Engineering, Loughborough University, Loughborough LE11 3TU (United Kingdom)

2017-01-05

Highlights: • Fe(III) shuttles electrons for enhanced reduction of Cr(VI) in MFCs. • The coulombic efficiency increases by 1.6 fold in the presence of Fe(III). • The reduction of Cr(VI) occurs via an indirect Fe(III) mediation mechanism. • Fe(III) decreases the diffusional resistances and the cathode overpotentials. - Abstract: The role of Fe(III) was investigated as an electron-shuttle mediator to enhance the reduction rate of the toxic heavy metal hexavalent chromium (Cr(VI)) in wastewaters, using microbial fuel cells (MFCs). The direct reduction of chromate (CrO{sub 4}{sup −}) and dichromate (Cr{sub 2}O{sub 7}{sup 2−}) anions in MFCs was hampered by the electrical repulsion between the negatively charged cathode and Cr(VI) functional groups. In contrast, in the presence of Fe(III), the conversion of Cr(VI) and the cathodic coulombic efficiency in the MFCs were 65.6% and 81.7%, respectively, 1.6 times and 1.4 folds as those recorded in the absence of Fe(III). Multiple analytical approaches, including linear sweep voltammetry, Tafel plot, cyclic voltammetry, electrochemical impedance spectroscopy and kinetic calculations demonstrated that the complete reduction of Cr(VI) occurred through an indirect mechanism mediated by Fe(III). The direct reduction of Cr(VI) with cathode electrons in the presence of Fe(III) was insignificant. Fe(III) played a critical role in decreasing both the diffusional resistance of Cr(VI) species and the overpotential for Cr(VI) reduction. This study demonstrated that the reduction of Cr(VI) in MFCs was effective in the presence of Fe(III), providing an alternative and environmentally benign approach for efficient remediation of Cr(VI) contaminated sites with simultaneous production of renewable energy.

Impact of Fe(III) as an effective electron-shuttle mediator for enhanced Cr(VI) reduction in microbial fuel cells: Reduction of diffusional resistances and cathode overpotentials

International Nuclear Information System (INIS)

Wang, Qiang; Huang, Liping; Pan, Yuzhen; Quan, Xie; Li Puma, Gianluca

2017-01-01

Highlights: • Fe(III) shuttles electrons for enhanced reduction of Cr(VI) in MFCs. • The coulombic efficiency increases by 1.6 fold in the presence of Fe(III). • The reduction of Cr(VI) occurs via an indirect Fe(III) mediation mechanism. • Fe(III) decreases the diffusional resistances and the cathode overpotentials. - Abstract: The role of Fe(III) was investigated as an electron-shuttle mediator to enhance the reduction rate of the toxic heavy metal hexavalent chromium (Cr(VI)) in wastewaters, using microbial fuel cells (MFCs). The direct reduction of chromate (CrO_4"−) and dichromate (Cr_2O_7"2"−) anions in MFCs was hampered by the electrical repulsion between the negatively charged cathode and Cr(VI) functional groups. In contrast, in the presence of Fe(III), the conversion of Cr(VI) and the cathodic coulombic efficiency in the MFCs were 65.6% and 81.7%, respectively, 1.6 times and 1.4 folds as those recorded in the absence of Fe(III). Multiple analytical approaches, including linear sweep voltammetry, Tafel plot, cyclic voltammetry, electrochemical impedance spectroscopy and kinetic calculations demonstrated that the complete reduction of Cr(VI) occurred through an indirect mechanism mediated by Fe(III). The direct reduction of Cr(VI) with cathode electrons in the presence of Fe(III) was insignificant. Fe(III) played a critical role in decreasing both the diffusional resistance of Cr(VI) species and the overpotential for Cr(VI) reduction. This study demonstrated that the reduction of Cr(VI) in MFCs was effective in the presence of Fe(III), providing an alternative and environmentally benign approach for efficient remediation of Cr(VI) contaminated sites with simultaneous production of renewable energy.

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)

AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT III, MAINTAINING THE FUEL SYSTEM--DETROIT DIESEL ENGINE.

Science.gov (United States)

Human Engineering Inst., Cleveland, OH.

THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM. TOPICS ARE (1) PURPOSE OF THE FUEL SYSTEM, (2) TRACING THE FUEL FLOW, (3) MINOR COMPONENTS OF THE FUEL SYSTEM, (4) MAINTENANCE TIPS, (5) CONSTRUCTION AND FUNCTION OF THE FUEL INJECTORS, AND (6)…

Diesel fueled ship propulsion fuel cell demonstration project

Energy Technology Data Exchange (ETDEWEB)

Kumm, W.H. [Arctic Energies Ltd., Severna Park, MD (United States)

1996-12-31

The paper describes the work underway to adapt a former US Navy diesel electric drive ship as a 2.4 Megawatt fuel cell powered, US Coast Guard operated, demonstrator. The Project will design the new configuration, and then remove the four 600 kW diesel electric generators and auxiliaries. It will design, build and install fourteen or more nominal 180 kW diesel fueled molten carbonate internal reforming direct fuel cells (DFCs). The USCG cutter VINDICATOR has been chosen. The adaptation will be carried out at the USCG shipyard at Curtis Bay, MD. A multi-agency (state and federal) cooperative project is now underway. The USCG prime contractor, AEL, is performing the work under a Phase III Small Business Innovation Research (SBIR) award. This follows their successful completion of Phases I and II under contract to the US Naval Sea Systems (NAVSEA) from 1989 through 1993 which successfully demonstrated the feasibility of diesel fueled DFCs. The demonstrated marine propulsion of a USCG cutter will lead to commercial, naval ship and submarine applications as well as on-land applications such as diesel fueled locomotives.

Complexation of trivalent actinides and lanthanides with hydrophilic N-donor ligands for Am(III)/Cm(III) and An(III)/Ln(III) separation; Komplexierung von trivalenten Actiniden und Lanthaniden mit hydrophilen N-Donorliganden zur Am(III)/Cm(III)- bzw. An(III)/Ln(III)-Trennung

Energy Technology Data Exchange (ETDEWEB)

Wagner, Christoph

2017-07-24

The implementation of actinide recycling processes is considered in several countries, aiming at the reduction of long-term radiotoxicity and heat load of used nuclear fuel. This requires the separation of the actinides from the fission and corrosion products. The separation of the trivalent actinides (An(III)) Am(III) and Cm(III), however, is complicated by the presence of the chemically similar fission lanthanides (Ln(III)). Hydrophilic N-donor ligands are employed as An(III) or Am(III) selective complexing agents in solvent extraction to strip An(III) or Am(III) from an organic phase loaded with An(III) and Ln(III). Though they exhibit excellent selectivity, the complexation chemistry of these ligands and the complexes formed during solvent extraction are not sufficiently characterized. In the present thesis the complexation of An(III) and Ln(III) with hydrophilic N-donor ligands is studied by time resolved laser fluorescence spectroscopy (TRLFS), UV/Vis, vibronic sideband spectroscopy and solvent extraction. TRLFS studies on the complexation of Cm(III) and Eu(III) with the Am(III) selective complexing agent SO{sub 3}-Ph-BTBP (tetrasodium 3,3{sup '},3'',3{sup '''}-([2,2{sup '}-bipyridine]-6,6{sup '}-diylbis(1,2,4-triazine-3,5,6-triyl)) tetrabenzenesulfonate) revealed the formation of [M(SO{sub 3}-Ph-BTBP){sub n}]{sup (4n-3)-} complexes (M = Cm(III), Eu(III); n = 1, 2). The conditional stability constants were determined in different media yielding two orders of magnitude larger β{sub 2}-values for the Cm(III) complexes, independently from the applied medium. A strong impact of ionic strength on the stability and stoichiometry of the formed complexes was identified, resulting from the stabilization of the pentaanionic [M(SO{sub 3}-Ph-BTBP){sub 2}]{sup 5-} complex with increasing ionic strength. Thermodynamic studies of Cm(III)-SO{sub 3}-Ph-BTBP complexation showed that the proton concentration of the applied medium impacts

IAEA activities on nuclear fuel

International Nuclear Information System (INIS)

Basak, U.

2011-01-01

In this paper a brief description and the main objectives of IAEA Programme B on Nuclear fuel cycle are given. The following Coordinated Research Projects: 1) FUel performance at high burn-up and in ageing plant by management and optimisation of WAter Chemistry Technologies (FUWAC ); 2) Near Term and Promising Long Term Options for Deployment of Thorium Based Nuclear Energy; 3) Fuel Modelling (FUMEX-III) are shortly described. The data collected by the IAEA Expert Group of Fuel Failures in Water Cooled Reactors including information about fuel failure cause for PWR (1994-2006) and failure mechanisms for BWR fuel (1994-2006) are shown. The just published Fuel Failure Handbook as well as preparation of a Monograph on Zirconium including an overview of Zirconium for nuclear applications are presented. The current projects in Sub-programme B2 - Power Reactor Fuel Engineering are also listed

40 CFR 80.535 - How are NRLM diesel fuel credits generated?

Science.gov (United States)

2010-07-01

... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive... the standards of § 80.510(a) or (b). V520 = The total volume of motor vehicle diesel fuel produced or... generated by both a foreign refiner and by an importer for the same motor vehicle diesel fuel. (iii) Credits...

Sulphur capture by co-firing sulphur containing fuels with biomass fuels - optimization

International Nuclear Information System (INIS)

Nordin, A.

1992-12-01

Previous results concerning co-firing of high sulphur fuels with biomass fuels have shown that a significant part of the sulphur can be absorbed in the ash by formation of harmless sulphates. The aim of this work has been to (i) determine the maximum reduction that can be obtained in a bench scaled fluidized bed (5 kW); (ii) determine which operating conditions will give maximum reduction; (iii) point out the importance and applicability of experimental designs and multivariate methods when optimizing combustion processes; (iv) determine if the degree of sulphur capture can be correlated to the degree of slagging, fouling or bed sintering; and (v) determine if further studies are desired. The following are some of the more important results obtained: - By co-firing peat with biomass, a total sulphur retention of 70 % can be obtained. By co-firing coal with energy-grass, the total SO 2 emissions can be reduced by 90 %. - Fuel feeding rate, amount of combustion air and the primary air ratio were the most important operating parameters for the reduction. Bed temperature and oxygen level seem to be the crucial physical parameters. - The NO emissions also decreased by the sulphur reducing measures. The CO emissions were relatively high (130 mg/MJ) compared to large scale facilities due to the small reactor and the small fluctuations in the fuel feeding rate. The SO 2 emissions could however be reduced without any increase in CO emissions. - When the reactor was fired with a grass, the bed sintered at a low temperature ( 2 SO 4 and KCl are formed no sintering problems were observed. (27 refs., 41 figs., 9 tabs., 3 appendices)

FFTF metal fuel pin sodium bond quality verification

International Nuclear Information System (INIS)

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

1988-12-01

The Fast Flux Test Facility (FFTF) Series III driver fuel design consists of U-10Zr fuel slugs contained in a ferritic alloy cladding. A liquid metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. This document discusses this verification

SIMMER-III parametric studies of fuel-steel mixing and radial mesh effects on power excursion in ESFR ULOF transients - 15033

International Nuclear Information System (INIS)

Chen, X.N.; Rineiski, A.; Gabrielli, F.; Andriolo, L.; Li, R.; Maschek, W.

2015-01-01

This paper deals with SIMMER-III once-through simulations of the first power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered initially in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is ∼ 2 dollars, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly coarse mesh treatment, where a simultaneous sodium boiling onset in all sub-assemblies belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped 'homogenization' reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the first power excursion after the sodium boiling: both effects delay the power excursion and significantly reduce the height of the power peaks in case of a ULOF

Separation by liquid-liquid extraction of actinides(III) from lanthanides(III) using new molecules: the picolinamides

International Nuclear Information System (INIS)

Cordier, P.Y.

1996-07-01

In the field of long-lived radionuclides separation from waste generated during spent fuel reprocessing, the picolinamides have been chosen as potential extractants for the selective extraction of actinides (III) from lanthanides (III). The first studies initiated on the most simple molecule of the picolinamide family, namely 2-pyridinecarboxamide, pointed out that in an aqueous media the complexation stability constant between this ligand and Am(III) is roughly 10 times higher than the ones corresponding to Ln(III). The synthesis of lipophilic derivatives of 2-pyridinecarboxamide leaded to extraction experiments. The extraction of metallic cation by lipophilic picolinamides, according to a solvatation mechanism, is strongly dependent on the nature of the amide function: a primary amide function (group I) leads to a good extraction; on the contrary, there is a decrease for secondary (group II) and tertiary (group III) amide functions. From a theoretical point of view, this work leads finally to the following conclusions: confirmation of the importance of the presence of soft donor atoms within the extractants (nitrogen in our case) for An(III)/Ln(III). Also, sensitivity of this soft donor atom regarding the protonation reaction; prevalence in our case of the affinity of the extractant for the metallic cation over the lipophilia of the extractant to ensure good distribution coefficients. The extraction and Am(III)/Ln(III) separation performances of the picolinamides from pertechnetic media leads to the design of a possible flowsheet for the reprocessing of high level liquid waste, with the new idea of an integrated technetium reflux. (author)

Analysis of excess reactivity of JOYO MK-III performance test core

International Nuclear Information System (INIS)

Maeda, Shigetaka; Yokoyama, Kenji

2003-10-01

JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

Interaction of Eu(III) and Cm(III) with mucin. A key component of the human mucosa

International Nuclear Information System (INIS)

Wilke, Claudia; Barkleit, Astrid

2017-01-01

To evaluate the potential health risks caused by the ingestion of lanthanides (Ln) and actinides (An), investigations into the chemical behavior of these metals in the human gastrointestinal tract are necessary. Mucin is an important part of the protective mucosa layer in the digestive system. We have recently reported that mucin interacts strongly with Eu(III) and Cm(III), representatives of Ln(III) and An(III), respectively, under in vivo conditions. In order to investigate the complexation behavior of this protein with Ln(III)/An(III), TRLFS measurements were performed on Eu(III)/Cm(III)-mucin solutions with different protein concentrations and at different pH. The results indicate the formation of at least two independent mucin species. At higher pH, the formation of hydroxide species was also observed.

Interaction of Eu(III) and Cm(III) with mucin. A key component of the human mucosa

Energy Technology Data Exchange (ETDEWEB)

Wilke, Claudia; Barkleit, Astrid [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Chemistry of the F-Elements

2017-06-01

To evaluate the potential health risks caused by the ingestion of lanthanides (Ln) and actinides (An), investigations into the chemical behavior of these metals in the human gastrointestinal tract are necessary. Mucin is an important part of the protective mucosa layer in the digestive system. We have recently reported that mucin interacts strongly with Eu(III) and Cm(III), representatives of Ln(III) and An(III), respectively, under in vivo conditions. In order to investigate the complexation behavior of this protein with Ln(III)/An(III), TRLFS measurements were performed on Eu(III)/Cm(III)-mucin solutions with different protein concentrations and at different pH. The results indicate the formation of at least two independent mucin species. At higher pH, the formation of hydroxide species was also observed.

Cubby : Multiscreen Desktop VR Part III

NARCIS (Netherlands)

Djajadiningrat, J.P.; Gribnau, M.W.

2000-01-01

In this month's final episode of our 'Cubby: Multiscreen Desktop VR' trilogy we explain how you read the InputSprocket driver from part II, how you use it as input for the cameras from part I and how you calibrate the input device so that it leads to the correct head position.

DUPIC fuel fabrication using spent PWR fuels at KAERI

Energy Technology Data Exchange (ETDEWEB)

Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

2000-12-01

This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter III, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

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.

Iron oxide redox chemistry and nuclear fuel disposal

International Nuclear Information System (INIS)

Jobe, D.J.; Lemire, R.J.; Taylor, P.

1997-04-01

Solubility and stability data for iron (III) oxides and aqueous Fe(II) and Fe(III) species are reviewed, and selected values are used to calculate potential-pH diagrams for the iron system at temperatures of 25 and 100 deg C, chloride activities {C1 - } = 10 -2 and 1 mol/kg, total carbonate activity {C T } = 10 -3 mol/kg, and iron(III) oxide/oxyhydroxide solubility products (25 deg C values) K sp = {Fe 3+ }{OH - } 3 = 10 -38.5 , 10 -40 and 10 -42 . The temperatures and anion concentrations bracket the range of conditions expected in a Canadian nuclear fuel waste disposal vault. The three solubility products represent a conservative upper limit, a most probable value, and a minimum credible value, respectively, for the iron oxides likely to be important in controlling redox conditions in a disposal vault for CANDU nuclear reactor fuel. Only in the first of these three cases do the calculated redox potentials significantly exceed values under which oxidative dissolution of the fuel may occur. (author)

SIMMER-III analytic thermophysical property model

International Nuclear Information System (INIS)

Morita, K; Tobita, Y.; Kondo, Sa.; Fischer, E.A.

1999-05-01

An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms are designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless steel, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-core materials. (author)

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.

Flow-induced vibration phenomenon in a Mark III TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, C K; Whittemore, W L; Kim, B S; Lee, J B; Blevins, R D; Burton, T E [Korea Atomic Energy Research Institute, Seoul (Korea, Republic of); General Atomic Company, San Diego, CA (United States)

1976-07-01

The Mark III TRIGA reactor with hexagonal fuel spacing is capable of operating at 2.0 MW. The Mark III at San Diego operated without core cooling problems or vibration at power levels up to 2.0 MW. All Mark III reactors have operated trouble-free up to 1.0 MW. The Mark III TRIGA in Korea was installed in 1972 and operated many months without trouble at 2.0 MW. During this period core changes including addition of new fuel were made. Eighteen months after startup, a coolant flow-induced vibration was observed for the first time at a power of 1.5 MW. A lengthy series of tests showed that it was not possible to establish a core configuration that permitted vibration-free operation for power levels in the range 1.5 - 2.0 MW. Observations during the tests confirmed that standing waves in the reactor tank water coupled the source within the core to the shield structure and surrounding building. Analysis of the data indicates strongly that the source of the vibration is the creation and collapse of bubbles with the core acting as a resonator. A substantially increased flow of coolant through the upper grid plate is expected to eliminate the vibration phenomenon and permit trouble-free operation at power up to 2.0 MW. In an attempt to seek a remedy, both GAC and KAERI have independently developed designs for upper grid plates. KAERI has constructed and installed an interim version of the standard grid plate which was calculated to provide 25% more coolant flow and mounted high so as to provide less restriction to flow around the upper fittings of the fuel elements. A substantial reduction in vibration was observed. No vibration was observed at any power up to 2.0 MW with cooling water at or below 20 C. A slight vibration at 1.8 MW occurred for higher cooling temperatures. The GAC grid plate design provides not only for increasing the flow area but also for streamlining the flow surfaces on the grid plate and possibly also on the top fittings of the fuel elements. It is

Flow-induced vibration phenomenon in a Mark III TRIGA reactor

International Nuclear Information System (INIS)

Lee, C.K.; Whittemore, W.L.; Kim, B.S.; Lee, J.B.; Blevins, R.D.; Burton, T.E.

1976-01-01

The Mark III TRIGA reactor with hexagonal fuel spacing is capable of operating at 2.0 MW. The Mark III at San Diego operated without core cooling problems or vibration at power levels up to 2.0 MW. All Mark III reactors have operated trouble-free up to 1.0 MW. The Mark III TRIGA in Korea was installed in 1972 and operated many months without trouble at 2.0 MW. During this period core changes including addition of new fuel were made. Eighteen months after startup, a coolant flow-induced vibration was observed for the first time at a power of 1.5 MW. A lengthy series of tests showed that it was not possible to establish a core configuration that permitted vibration-free operation for power levels in the range 1.5 - 2.0 MW. Observations during the tests confirmed that standing waves in the reactor tank water coupled the source within the core to the shield structure and surrounding building. Analysis of the data indicates strongly that the source of the vibration is the creation and collapse of bubbles with the core acting as a resonator. A substantially increased flow of coolant through the upper grid plate is expected to eliminate the vibration phenomenon and permit trouble-free operation at power up to 2.0 MW. In an attempt to seek a remedy, both GAC and KAERI have independently developed designs for upper grid plates. KAERI has constructed and installed an interim version of the standard grid plate which was calculated to provide 25% more coolant flow and mounted high so as to provide less restriction to flow around the upper fittings of the fuel elements. A substantial reduction in vibration was observed. No vibration was observed at any power up to 2.0 MW with cooling water at or below 20 C. A slight vibration at 1.8 MW occurred for higher cooling temperatures. The GAC grid plate design provides not only for increasing the flow area but also for streamlining the flow surfaces on the grid plate and possibly also on the top fittings of the fuel elements. It is

Fuel for the next Brazilian nuclear power plants

International Nuclear Information System (INIS)

Lameiras, Fernando S.; Faeda, Kelly Cristina Ferreira

2009-01-01

The conclusion of the Angra III nuclear power plant ends a cycle of the nuclear energy in Brazil that started about forty years ago. Nowadays the country is planning the installation of 4 GWe to 8 GWe of nuclear power up to the year 2030. The nuclear reactors considered for this new cycle should take into account the current technologic development and environment of the nuclear market. They certainly will have significant differences in relation to the Angra I, II, and III reactors. Important impacts may result on the nuclear fuel production chain, e. g., case high temperature reactors were chosen, which can deliver electricity and heat. The differences between the fuels of the candidate reactors after Angra III are analyzed and development lines are suggested to minimize these impacts. (author)

Fuel Rod Consolidation Project: Phase 2, Final report: Volume 5, Operations and maintenance manual

International Nuclear Information System (INIS)

1988-01-01

The purpose of this manual is to describe the function, installation, operation and maintenance of the Fuel Rod Consolidation System. This Document is preliminary and must be updated to incorporate any modifications to the mechanical and electrical systems that are performed during construction. Any changes and specific references related to the software requirements will be provided as the software is developed in Phase III. Setpoints related to equipment positions as a function of resolver and position transducer readings will also be provided in Phase III. References such as vendor supplied Operating and Maintenance Manuals for vendor components and assemblies are not available until a receipt of a purchase order. These references will become an integral part of this manual during the construction phase

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.

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

1992-01-01

The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

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.

PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit B: Financial Management. Research and Development Series No. 194 C-2.

Science.gov (United States)

Ohio State Univ., Columbus. National Center for Research in Vocational Education.

This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit B focuses on good financial management…

PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit H: Business Protection. Research and Development Series No. 194 C-8.

Science.gov (United States)

Ohio State Univ., Columbus. National Center for Research in Vocational Education.

This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in part III is operating a business. Unit H focuses on business protection. It…

PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit G: Community Relations. Research and Development Series No. 194 C-7.

Science.gov (United States)

Ohio State Univ., Columbus. National Center for Research in Vocational Education.

This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups of vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit G focuses on community relations. It…

PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit D: Marketing Management. Research and Development Series No. 194 C-4.

Science.gov (United States)

Ohio State Univ., Columbus. National Center for Research in Vocational Education.

This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit D focuses on market management. It…

Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

FUMEX-III Cases by START-3. Final Report

International Nuclear Information System (INIS)

Novikov, Vladimir; Bogatyr, Sergey; Kuznetsov, Vladimir; Chulkin, Dmitriy

2013-01-01

This document covers ''SUPER-RAMP'' and ''AREVA High-Burnup Idealised case'', cases of FUMEX-III project that were calculated by means of START-3 code. SUPER-RAMP PK-2 rodlets were chosen for calculation because they were subjected to strong power ramp and did not fail, because the main objective of the present calculations was to check the models of fission gas release and gaseous swelling of START-3 in the conditions of the power ramps (we did not intend to get the limit stresses for Zry-4 fuel rods). AREVA high burnup (PRIORITY CASE) case was chosen to check the applicability of the START-3 code to high-burnup PWR fuel rods behavior. Also, in order to check the stability and quality of the changes made into the code during the course of FUMEX-III participation, several simplified cases of FUMEX-II were recalculated. FUMEX-III SUPER-RAMP exercise included up to 49 kW/m power ramps of fuel rods with burnup up to 45 MWd/kgU. START-3 calculations of this exercise are in reasonable agreement with the experiment. AREVA high burnup priority case (∼81.5 MWd/kgU) was calculated and overprediction of the FGR was found. This overprediction is related with the conservative overestimation of rim-structure capability to retain fission gas and can be adjusted implementing a simple correction of this property. To verify the changes made into the code (oxide layer growth model and Zry-4 radiation growth), the corrected version of the code was used to recalculate the FUMEX-II simplified cases 27.2 and satisfactory results were obtained. (author)

17 CFR Table III to Subpart E of... - Civil Monetary Penalty Inflation Adjustments

Science.gov (United States)

2010-04-01

... 17 Commodity and Securities Exchanges 2 2010-04-01 2010-04-01 false Civil Monetary Penalty Inflation Adjustments III Table III to Subpart E of Part 201 Commodity and Securities Exchanges SECURITIES..., Table III Table III to Subpart E of Part 201—Civil Monetary Penalty Inflation Adjustments U.S. Code...

Comparison of SAS3A and MELT-III predictions for a transient overpower hypothetical accident

International Nuclear Information System (INIS)

Wilburn, N.P.

1976-01-01

A comparison is made of the predictions of the two major codes SAS3A and MELT-III for the hypothetical unprotected transient overpower accident in the FFTF. The predictions of temperatures, fuel restructuring, fuel melting, reactivity feedbacks, and core power are compared

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)

PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit E: Successful Selling. Research and Development Series No. 194 C-5.

Science.gov (United States)

Ohio State Univ., Columbus. National Center for Research in Vocational Education.

This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit E focuses on personal (face-to-face)…

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

Combustion chemistry and flame structure of furan group biofuels using molecular-beam mass spectrometry and gas chromatography - Part III: 2,5-Dimethylfuran.

Science.gov (United States)

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

2014-03-01

This work is the third part of a study focusing on the combustion chemistry and flame structure of furan and selected alkylated derivatives, i.e. furan in Part I, 2-methylfuran (MF) in Part II, and 2,5-dimethylfuran (DMF) in the present work. Two premixed low-pressure (20 and 40 mbar) flat argon-diluted (50%) flames of DMF were studied with electron-ionization molecular-beam mass spectrometry (EI-MBMS) and gas chromatography (GC) under two equivalence ratios (φ=1.0 and 1.7). Mole fractions of reactants, products, and stable and radical intermediates were measured as a function of the distance to the burner. Kinetic modeling was performed using a reaction mechanism that was further developed in the present series, including Part I and Part II. A reasonable agreement between the present experimental results and the simulation is observed. The main reaction pathways of DMF consumption were derived from a reaction flow analysis. Also, a comparison of the key features for the three flames is presented, as well as a comparison between these flames of furanic compounds and those of other fuels. An a priori surprising ability of DMF to form soot precursors (e.g. 1,3-cyclopentadiene or benzene) compared to less substituted furans and to other fuels has been experimentally observed and is well explained in the model.

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.)

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Fuel reprocessing and waste management

International Nuclear Information System (INIS)

Philippone, R.L.; Kaiser, R.A.

1989-01-01

Because of different economic, social and political factors, there has been a tendency to compartmentalize the commercial nuclear power industry into separate power and fuel cycle operations to a greater degree in some countries compared to other countries. The purpose of this paper is to describe how actions in one part of the industry can affect the other parts and recommend an overall systems engineering approach which incorporates more cooperation and coordination between individual parts of the fuel cycle. Descriptions are given of the fuel cycle segments and examples are presented of how a systems engineering approach has benefitted the fuel cycle. Descriptions of fuel reprocessing methods and the waste forms generated are given. Illustrations are presented describing how reprocessing options affect waste management operations and how waste management decisions affect reprocessing

Precipitation of plutonium (III) oxalate and calcination to plutonium oxide

International Nuclear Information System (INIS)

Esteban, A.; Orosco, E.H.; Cassaniti, P.; Greco, L.; Adelfang, P.

1989-01-01

The plutonium based fuel fabrication requires the conversion of the plutonium nitrate solution from nuclear fuel reprocessing into pure PuO2. The conversion method based on the precipitation of plutonium (III) oxalate and subsequent calcination has been studied in detail. In this procedure, plutonium (III) oxalate is precipitated, at room temperature, by the slow addition of 1M oxalic acid to the feed solution, containing from 5-100 g/l of plutonium in 1M nitric acid. Before precipitation, the plutonium is adjusted to trivalent state by addition of 1M ascorbic acid in the presence of an oxidation inhibitor such as hydrazine. Finally, the precipitate is calcinated at 700 deg C to obtain PuO2. A flowsheet is proposed in this paper including: a) A study about the conditions to adjust the plutonium valence. b) Solubility data of plutonium (III) oxalate and measurements of plutonium losses to the filtrate and wash solution. c) Characterization of the obtained products. Plutonium (III) oxalate has several potential advantages over similar conversion processes. These include: 1) Formation of small particle sizes powder with good pellets fabrication characteristics. 2) The process is rather insensitive to most process variables, except nitric acid concentration. 3) Ambient temperature operations. 4) The losses of plutonium to the filtrate are less than in other conversion processes. (Author) [es

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.

The history, genotoxicity and carcinogenicity of carbon-based fuels and their emissions: part 4 - alternative fuels.

Science.gov (United States)

Claxton, Larry D

2015-01-01

Much progress has been made in reducing the pollutants emitted from various combustors (including diesel engines and power plants) by the use of alternative fuels; however, much more progress is needed. Not only must researchers improve fuels and combustors, but also there is a need to improve the toxicology testing and analytical chemistry methods associated with these complex mixtures. Emissions from many alternative carbonaceous fuels are mutagenic and carcinogenic. Depending on their source and derivation, alternative carbonaceous fuels before combustion may or may not be genotoxic; however, in order to know their genotoxicity, appropriate chemical analysis and/or bioassay must be performed. Newly developed fuels and combustors must be tested to determine if they provide a public health advantage over existing technologies - including what tradeoffs can be expected (e.g., decreasing levels of PAHs versus increasing levels of NOx and possibly nitroarenes in ambient air). Another need is to improve exposure estimations which presently are a weak link in doing risk analyses. Copyright © 2014 Elsevier B.V. All rights reserved.

PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit F: Managing Human Resources. Research and Development Series No. 194 C-6.

Science.gov (United States)

Ohio State Univ., Columbus. National Center for Research in Vocational Education.

This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-level colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit F focuses on proper management of human…

PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit A: Managing the Business. Research and Development Series No. 194 C-1.

Science.gov (United States)

Ohio State Univ., Columbus. National Center for Research in Vocational Education.

This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit A focuses on the management process. It…

Plutonium fuel program

International Nuclear Information System (INIS)

1979-01-01

The work of the Project-Fuel Development reached the apex of its current programme during the course of the year. Notable success was recorded in the area of irradiation testing with the completion of the examination of the MFBS-7 irradiation. The irradiation group also prepared the seventh Filos experiment and this, as well as the DIDO-III test, began irradiation at the end of the year. Consideration was given and plans prepared for a revised pin filling line for bundle tests. Work also began on the conceptual design study for a pilot production line having a nominal capacity of 500 kg fuel per year. (Auth.)

Part 6. Internationalization and collocation of FBR fuel cycle facilities

International Nuclear Information System (INIS)

Stevenson, M.G.; Abramson, P.B.; LeSage, L.G.

1980-01-01

This report examines some of the non-proliferation, technical, and institutional aspects of internationalization and/or collocation of major facilities of the Fast Breeder Reactor (FBR) fuel cycle. The national incentives and disincentives for establishment of FBR Fuel Cycle Centers are enumerated. The technical, legal, and administrative considerations in determining the feasibility of FBR Fuel Cycle Centers are addressed by making comparisons with Light Water Reactor (LWR) centers which have been studied in detail by the IAEA and UNSRC

A comparison of FEMAXI-III code calculations with irradiation experiments

International Nuclear Information System (INIS)

Ito, K.; Sogame, M.; Ichikawa, M.; Nakajima, T.

1981-01-01

The FEMAXI-III code calculations were compared with in-pile diameter measurements in the Halden Boiling Water Reactor, in order to check the ability to analyse the pellet-cladding mechanical interaction. The results showed generally good agreement between calculations and measurements. The Studsvik INTER-RAMP Experiments were also analysed to examine the predictability of fuel rod failures. Good agreement was obtained between calculated and measured fission gas x release. The threshold stress to cause failure was estimated by means of FEMAXI-III. (author)

Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, IV Deo, Promena izotopnog sastava goriva

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-07-15

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{sup 235} chain, analytical expressions for the concentrations of U{sup 235}, U{sup 236} and Np{sup 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.

A method of failed fuel detection

International Nuclear Information System (INIS)

Uchida, Shunsuke; Utamura, Motoaki; Urata, Megumu.

1976-01-01

Object: To keep the coolant fed to a fuel assembly at a level below the temperature of existing coolant to detect a failed fuel with high accuracy without using a heater. Structure: When a coolant in a coolant pool disposed at the upper part of a reactor container is fed by a coolant feed system into a fuel assembly through a cap to fill therewith and exchange while forming a boundary layer between said coolant and the existing coolant, the temperature distribution of the feed coolant is heated by fuel rods so that the upper part is low whereas the lower part is high. Then, the lower coolant is upwardly moved by the agitating action and fission products leaked through a failed opening at the lower part of the fuel assembly and easily extracted by the sampling system. (Yoshino, Y.)

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)

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

AUTOMOTIVE DIESEL MAINTENANCE L. UNIT XII, PART I--MAINTAINING THE FUEL SYSTEM (PART II), CUMMINS DIESEL ENGINE, PART II--UNIT INSTALLATION (ENGINE).

Science.gov (United States)

Human Engineering Inst., Cleveland, OH.

THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM AND THE PROCEDURES FOR DIESEL ENGINE INSTALLATION. TOPICS ARE FUEL FLOW CHARACTERISTICS, PTG FUEL PUMP, PREPARATION FOR INSTALLATION, AND INSTALLING ENGINE. THE MODULE CONSISTS OF A SELF-INSTRUCTIONAL BRANCH…

Thermomechanical analysis of nuclear fuel elements

International Nuclear Information System (INIS)

Hernandez L, H.

1997-01-01

This work presents development of a code to obtain the thermomechanical analysis of fuel rods in the fuel assemblies inserted in the core of BWR reactors. The code uses experimental correlations developed in several laboratories. The development of the code is divided in two parts: a) the thermal part and b) the mechanical part, extending both the fuel and the cladding materials. The thermal part consists of finding the radial distribution of temperatures in the pellet, from the fuel centerline up to the coolant, along the total active length, considering one and two phase flow in the coolant, as a result of the pressure drop in the system. The mechanical part analyzes the effects of temperature gradients, pressure and irradiation, to which the fuel rod is subjected. The strains produced by swelling, creep and thermal stress in the fuel material are analyzed. In the same way the strains in the cladding are analyzed, considering the effects produced by the pressure exerted on the cladding by pellet swelling, by the pressure caused by fission gas release toward the cavities, and by the strain produced on the cladding by the pressure changes of the system. (Author)

Nuclear Fuel Cycle Evaluation and Real Options

Directory of Open Access Journals (Sweden)

L. Havlíček

2008-01-01

Full Text Available The first part of this paper describes the nuclear fuel cycle. It is divided into three parts. The first part, called Front-End, covers all activities connected with fuel procurement and fabrication. The middle part of the cycle includes fuel reload design activities and the operation of the fuel in the reactor. Back-End comprises all activities ensuring safe separation of spent fuel and radioactive waste from the environment. The individual stages of the fuel cycle are strongly interrelated. Overall economic optimization is very difficult. Generally, NPV is used for an economic evaluation in the nuclear fuel cycle. However the high volatility of uranium prices in the Front-End, and the large uncertainty of both economic and technical parameters in the Back-End, make the use of NPV difficult. The real option method is able to evaluate the value added by flexibility of decision making by a company under conditions of uncertainty. The possibility of applying this method to the nuclear fuel cycle evaluation is studied. 

Mitigating fuel handling situations during station blackout in TAPP-3 and

International Nuclear Information System (INIS)

Chugh, V.K.; Roy, Shibaji; Gupta, H.; Inder Jit

2002-01-01

Full text: On power refueling is one of the important features of PHWRs. fuelling machine (FM) Head becomes part of the reactor pressure boundary during refueling operations. Hot irradiated (spent) fuel bundles are received in the FM Head from the Reactor and transferred to spent fuel storage bay (SFSB). These bundles pass through various fuel handling (FH) Equipment under submerged condition except during the dry transfer operation. Situations of station blackout (SBO) i.e. postulated simultaneous failure of Class III and Class IV electric power, could persist for a long period, during on-reactor or off-reactor FH operations, with the spent fuel bundles being any where in the system between the reactor and SFSB. The cooling provisions for the spent fuel bundles vary depending upon the stage of operation. During SBO, it becomes difficult to maintain cooling to these fuel bundles due to the limited availability of Class II power and instrument air. However, cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like stay-put, gravity- fill, D 2 O-steaming etc. for cooling the bundles. Various scenarios have been identified for cooling provisions of the bundles in the system. The paper also describes consequences like loss of D 2 O inventory, rise in ambient temperature and pressure and tritium build-up in Reactor Building, emanating from these cooling schemes

Failure analysis of carbide fuels under transient overpower (TOP) conditions

International Nuclear Information System (INIS)

Nguyen, D.H.

1980-06-01

The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin

Contribution of the Politecnico di Milano to the FUMEX-III Project

International Nuclear Information System (INIS)

Luzzi, Lelio; Pastore, Giovanni; Van Uffelen, Paul

2013-01-01

In nuclear reactors, irradiation continuously alter the thermal, mechanical, and chemical properties of nuclear fuels (Olander 1976). To assure the safe and economic operation of the nuclear fuel rods in all the operation conditions, there is a need for fuel characterization and optimization through an integrated theoretical, experimental, and computational approach. The aim of computational fuel modelling is to predict the changes in properties and evaluate the thermo-mechanical behaviour of the fuel rods during the life in the reactor. For this purpose, increasingly complex fuel performance codes are developed, which include physical models of the processes taking place in the fuel rods during irradiation (Aybar and Ortego 2005). In this framework, international benchmark exercises on fuel modelling are of high importance for the development of fuel performance codes, since they provide the possibility for cross-comparison and complementary validation of a large number of codes involved. Three such exercises were organized during the last 3 decades: D-COM in the mid 80's (Misfeldt 1983), and the Coordinated Research Projects (CRPs) FUMEX-I (1993-1996) (Chantoin et al. 1997) and FUMEX-II (2002- 2006) (Killeen et al. 2007, IAEA 2011). In extending the previous CRPs on the subject of improving the predictive capabilities of fuel performance codes for extended burn-up and transient conditions, the focus of the CRP FUMEX-III (2008-2012) is on the topics of fission gas release, pellet-cladding interaction (PCI) and dimensional changes (Killeen et al. 2009). The TRANSURANUS fuel performance code (Lassmann 1992, Lassmann 2001) is presently available at the Politecnico di Milano (POLIMI). Based on the assumption of axial-symmetric cylindrical rod and the superposition of a one-dimensional radial and axial description (11/2D approach), the mechanical-mathematical framework of TRANSURANUS allows to analyze, at reasonable computer cost, the integral fuel rod during a

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

Fuel processing

International Nuclear Information System (INIS)

Allardice, R.H.

1990-01-01

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

Combustion chemistry and flame structure of furan group biofuels using molecular-beam mass spectrometry and gas chromatography – Part III: 2,5-Dimethylfuran

Science.gov (United States)

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

2013-01-01

This work is the third part of a study focusing on the combustion chemistry and flame structure of furan and selected alkylated derivatives, i.e. furan in Part I, 2-methylfuran (MF) in Part II, and 2,5-dimethylfuran (DMF) in the present work. Two premixed low-pressure (20 and 40 mbar) flat argon-diluted (50%) flames of DMF were studied with electron-ionization molecular-beam mass spectrometry (EI-MBMS) and gas chromatography (GC) under two equivalence ratios (φ=1.0 and 1.7). Mole fractions of reactants, products, and stable and radical intermediates were measured as a function of the distance to the burner. Kinetic modeling was performed using a reaction mechanism that was further developed in the present series, including Part I and Part II. A reasonable agreement between the present experimental results and the simulation is observed. The main reaction pathways of DMF consumption were derived from a reaction flow analysis. Also, a comparison of the key features for the three flames is presented, as well as a comparison between these flames of furanic compounds and those of other fuels. An a priori surprising ability of DMF to form soot precursors (e.g. 1,3-cyclopentadiene or benzene) compared to less substituted furans and to other fuels has been experimentally observed and is well explained in the model. PMID:24518851

Convective parameters in fuel elements for research nuclear reactors

International Nuclear Information System (INIS)

Lopez Martinez, C.D.

1992-01-01

The study of a prototype for the simulation of fuel elements for research nuclear reactors by natural convection in water is presented in this paper. This project is carry out in the thermofluids laboratory of National Institute of Nuclear Research. The fuel prototype has already been test for natural convection in air, and the first results in water are presented in this work. In chapter I, a general description of Triga Mark III is made, paying special atention to fuel-moderator components. In chapter II and III an approach to convection subject in its global aspects is made, since the intention is to give a general idea of the events occuring around fuel elements in a nuclear reactor. In chapter II, where an emphasis on forced convection is made, some basic concepts for forced convection as well as for natural convection are included. The subject of flow through cylinders is annotated only as a comparative reference with natural convection in vertical cylinders, noting the difference between used correlations and the involved variables. In chapter III a compilation of correlation found in the bibliography about natural convection in vertical cylinders is presented, since its geometry is the more suitable in the analysis of a fuel rod. Finally, in chapter IV performed experiments in the test bench are detailed, and the results are presented in form of tables and graphs, showing the used equations for the calculations and the restrictions used in each case. For the analysis of the prototypes used in the test bench, a constant and uniform flow of heat in the whole length of the fuel rod is considered. At the end of this chapter, the work conclusions and a brief explanation of the results are presented (Author)

Nitrato-complexes of Y(III), La(III), Ce(III), Pr(III), Nd(III), Sm(III), Gd(III), Tb(III), Dy(III) and Ho(III) with 2-(2'-pyridyl) benzimidazole

Energy Technology Data Exchange (ETDEWEB)

Mishra, A; Singh, M P; Singh, V K

1982-05-01

The nitrato-complexes, (Y(PyBzH)/sub 2/(NO/sub 3/)/sub 2/)NO/sub 3/.H/sub 2/O and Nd, Sm, Gd, Tb, Dy, Ho ; n=1-3, m=0-0.5 ; PyBzh=2-(2 -pyridyl)benzimidazole) are formed on interaction of the ligand with metal nitrates in ethanol. The electrical conductance values (116-129 ohm/sup -1/cm/sup 2/mol/sup -1/) suggest 1:1 electrolyte-nature of the complexes. Magnetic moment values of Ce(2.53 B.M.), Pr(3.62 B.M.), Nd(3.52 B.M.), Sm(1.70 B.M.), Gd(8.06 B.M.), Tb(9.44 B.M.), Dy(10.56 B.M.) and Ho(10.51 B.M.) in the complexes confirm the positive state of the metals. Infrared evidences are obtained for the existance of both coordinated (C/sub 2/v) and uncoordinated (D/sub 3/h) nitrate groups. Electronic absorption spectra of Pr(III)-, Nd(III)-, Sm(III)-, Tb(III)-, Dy(III)- and Ho(III)-complexes have been analysed in the light of LSJ terms.

Nitrato-complexes of Y(III), La(III), Ce(III), Pr(III), Nd(III), Sm(III), Gd(III), Tb(III), Dy(III) and Ho(III) with 2-(2'-pyridyl) benzimidazole

International Nuclear Information System (INIS)

Mishra, A.; Singh, M.P.; Singh, V.K.

1982-01-01

The nitrato-complexes, [Y(PyBzH) 2 (NO 3 ) 2 ]NO 3 .H 2 O and Nd, Sm, Gd, Tb, Dy, Ho ; n=1-3, m=0-0.5 ; PyBzh=2-(2 -pyridyl)benzimidazole] are formed on interaction of the ligand with metal nitrates in ethanol. The electrical conductance values (116-129 ohm -1 cm 2 mol -1 ) suggest 1:1 electrolyte-nature of the complexes. Magnetic moment values of Ce(2.53 B.M.), Pr(3.62 B.M.), Nd(3.52 B.M.), Sm(1.70 B.M.), Gd(8.06 B.M.), Tb(9.44 B.M.), Dy(10.56 B.M.) and Ho(10.51 B.M.) in the complexes confirm the terpositive state of the metals. Infrared evidences are obtained for the existance of both coordinated (C 2 v) and uncoordinated (D 3 h) nitrate groups. Electronic absorption spectra of Pr(III)-, Nd(III)-, Sm(III)-, Tb(III)-, Dy(III)- and Ho(III)-complexes have been analysed in the light of LSJ terms. (author)

Validation and Improvement of the FEMAXI-JNES Code by Using PIE Data at Extended Burnup. Final Report for FUMEX-III

International Nuclear Information System (INIS)

Hirose, Tsutomu; Miura, Hiromichi; Kitamura, Toshiya; Kamimura, Katsuichiro

2013-01-01

Japan Nuclear Energy Safety Organization (JNES) has participated in the IAEA FUMEX-III Coordinated Research Project (CRP) on the Improvement of Computer Codes Used for Fuel Behaviour Simulation for the following purpose. 1. Cooperate between member states and exchange information and expertise for understanding of fuel modelling and improvement 2. Develop and improve the FEMAXI-JNES code as an audit code for Japanese safety licensing review of fuel rod design, especially, - High burnup fuel - MOX fuel 3. Set the standard models for the FEMAXI-JNES code to provide best-estimate predictions of the thermal and mechanical performance of LWR fuel rod This is the JNES's final report for the FUMEX-III CRP. During the period of the CRP, JNES has modified pellet swelling and fission gas release models, and demonstrated the predictive capability relative to fuel centerline temperature, fission gas release, fuel rod internal gas pressure, cladding diametral deformation and cladding elongation by comparisons of integral code predictions of these parameters to experimental (measured) data from OECD/NEA IFPE database. (author)

Applications for fueling of Forsmark-3 and Oskarshamn III

International Nuclear Information System (INIS)

1984-01-01

The method for handling and final disposal of spent nuclear fuel outlined in the report KBS-3 has been found acceptable in relation to conventional and radiation safety. The main institutions agree on this, IAEA and other international institutions deem the method over-safe. The KBS-3 study has left some problems unsolved. Further research projects have been identified. (Aa)

Performance and fuel conversion efficiency of a spark ignition engine fueled with iso-butanol

International Nuclear Information System (INIS)

Irimescu, Adrian

2012-01-01

Highlights: ► Iso-butanol use in a port injection spark ignition engine. ► Fuel conversion efficiency calculated based on chassis dynamometer measurements. ► Combined study of engine efficiency and air–fuel mixture temperature. ► Excellent running characteristics with minor fuel system modifications. ► Up to 11% relative drop in part load efficiency due to incomplete fuel vaporization. -- Abstract: Alcohols are increasingly used as fuels for spark ignition engines. While ethanol is most commonly used, long chain alcohols such as butanol feature several advantages like increased heating value and reduced corrosive action. This study investigated the effect of fueling a port injection engine with iso-butanol, as compared to gasoline operation. Performance levels were maintained within the same limits as with the fossil fuel without modifications to any engine component. An additional electronic module was used for increasing fuel flow by extending the injection time. Fuel conversion efficiency decreased when the engine was fueled with iso-butanol by up to 9% at full load and by up to 11% at part load, calculated as relative values. Incomplete fuel evaporation was identified as the factor most likely to cause the drop in engine efficiency.

Spectroscopic investigations on the complexation of Cm(III) and Eu(III) with organic model ligands and their binding mode in human urine (in vitro); Spektroskopische Untersuchungen zur Komplexbildung von Cm(III) und Eu(III) mit organischen Modellliganden sowie ihrer chemischen Bindungsform in menschlichem Urin (in vitro)

Energy Technology Data Exchange (ETDEWEB)

Heller, Anne

2011-10-26

In case of incorporation, trivalent actinides (An(III)) and lanthanides (Ln(III)) pose a serious health risk to humans. An(III) are artificial, highly radioactive elements which are mainly produced during the nuclear fuel cycle in nuclear power plants. Via hazardous accidents or nonprofessional storage of radioactive waste, they can be released in the environment and enter the human food chain. In contrast, Ln(III) are nonradioactive, naturally occurring elements with multiple applications in technique and medicine. Consequently it is possible that humans get in contact and incorporate both, An(III) and Ln(III). Therefore, it is of particular importance to elucidate the behaviour of these elements in the human body. While macroscopic processes such as distribution, accumulation and excretion are studied quite well, knowledge about the chemical binding form (speciation) of An(III) and Ln(III) in various body fluids is still sparse. In the present work, for the first time, the speciation of Cm(III) and Eu(III) in natural human urine (in vitro) has been investigated spectroscopically and the formed complex identified. For this purpose, also basic investigations on the complex formation of Cm(III) and Eu(III) in synthetic model urine as well as with the urinary relevant, organic model ligands urea, alanine, phenylalanine, threonine and citrate have been performed and the previously unknown complex stability constants determined. Finally, all experimental results were compared to literature data and predictions calculated by thermodynamic modelling. Since both, Cm(III) and Eu(III), exhibit unique luminescence properties, particularly the suitability of time-resolved laser-induced fluorescence spectroscopy (TRLFS) could be demonstrated as a method to investigate these metal ions in untreated, complex biofluids. The results of this work provide new scientific findings on the biochemical reactions of An(III) and Ln(III) in human body fluids on a molecular scale and

Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B.

2003-01-01

The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

Providing for energy efficiency in homes and small buildings. Part III. Determining which practices are most effective and installing materials

Energy Technology Data Exchange (ETDEWEB)

None

1980-06-01

The training program is designed to educate students and individuals in the importance of conserving energy and to provide for developing skills needed in the application of energy-saving techniques that result in energy-efficient buildings. A teacher guide and student workbook are available to supplement the basic manual. Subjects covered in Part III are: determining which practices are most efficient and economical; installing energy-saving materials; and improving efficiency of equipment.

Thermohydraulic analysis of assemblies containing up to 2/7 fuel rods

International Nuclear Information System (INIS)

Ferreira, W.J.; Luz, M.

1985-01-01

The COBRA IV-I computer code was tested using data from the Fast Flux Test Facility. Then this code was applied to the analysis of fuel assemblies from the Binary Breeder Reactor. Previously this analysis was carried out using the COBRA III-C code which allows only for the calculations of fuel assemblies having seven fuel pins. The COBRA IV-I permits the calculation of fuel assemblies containing up to 217 fuel pins and the inclusion of blanket and shielding effects. (F.E.) [pt

New type fuel exchange system

International Nuclear Information System (INIS)

Meshii, Toshio; Maita, Yasushi; Hirota, Koichi; Kamishima, Yoshio.

1988-01-01

When the reduction of the construction cost of FBRs is considered from the standpoint of the machinery and equipment, to make the size small and to heighten the efficiency are the assigned mission. In order to make a reactor vessel small, it is indispensable to decrease the size of the equipment for fuel exchange installed on the upper part of a core. Mitsubishi Heavy Industries Ltd. carried out the research on the development of a new type fuel exchange system. As for the fuel exchange system for FBRs, it is necessary to change the mode of fuel exchange from that of LWRs, such as handling in the presence of chemically active sodium and inert argon atmosphere covering it and handling under heavy shielding against high radiation. The fuel exchange system for FBRs is composed of a fuel exchanger which inserts, pulls out and transfers fuel and rotary plugs. The mechanism adopted for the new type fuel exchange system that Mitsubishi is developing is explained. The feasibility of the mechanism on the upper part of a core was investigated by water flow test, vibration test and buckling test. The design of the mechanism on the upper part of the core of a demonstration FBR was examined, and the new type fuel exchange system was sufficiently applicable. (Kako, I.)

Analysis of the microturbine combustion chamber by using the CHEMKIN III computer code; Analise da camara de combustao de microturbinas empregando-se o codigo computacional CHEMKIN III

Energy Technology Data Exchange (ETDEWEB)

Madela, Vinicius Zacarias; Pauliny, Luis F. de A.; Veras, Carlos A. Gurgel [Brasilia Univ., DF (Brazil). Dept. de Engenharia Mecanica]. E-mail: gurgel@enm.unb.br; Costa, Fernando de S. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Combustao e Propulsao]. E-mail: fernando@cptec.inpe.br

2000-07-01

This work presents the results obtained with the simulation of multi fuel micro turbines combustion chambers. In particular, the predictions for the methane and Diesel burning are presented. The appropriate routines of the CHEMKIN III computer code were used.

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

2015-12-15

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

Acuity and case management: a healthy dose of outcomes, part III.

Science.gov (United States)

Huber, Diane L; Craig, Kathy

2007-01-01

This is the third of a 3-part series presenting 2 effective applications--acuity and dosage--that describe how the business case for case management (CM) can be made. In Part I, dosage and acuity concepts were explained as client need-severity, CM intervention-intensity, and CM activity-dose prescribed by amount, frequency, duration, and breadth of activities. Concepts were presented that related the practice of CM to the use of evidence-based practice (EBP), knowledge, and methods and the development of instruments that measure and score pivotal CM actions. Part I also featured a specific exemplar, the CM Acuity Tool, and described how to use acuity to identify and score the complexity of a CM case. Part II further explained dosage and 2 acuity instruments, the Acuity Tool and AccuDiff. Part III presents linkage to EBP and practical applications. The information contained in the 3-part series applies to all CM practice settings and contains ideas and recommendations useful to CM generalists, specialists, supervisors, and business and outcomes managers. The Acuity Tools Project was developed from frontline CM practice in one large, national telephonic CM company. Dosage: A literature search failed to find research into dosage of a behavioral intervention. The Huber-Hall model was developed and tested in a longitudinal study of CM models in substance abuse treatment and reported in the literature. Acuity: A structured literature search and needs assessment launched the development of the suite of acuity tools. A gap analysis identified that an instrument to assign and measure case acuity specific to CM activities was needed. Clinical experts, quality specialists, and business analysts (n = 7) monitored the development and testing of the tools, acuity concepts, scores, differentials, and their operating principles and evaluated the validity of the acuity tools' content related to CM activities. During the pilot phase of development, interrater reliability testing of

A Structural Molar Volume Model for Oxide Melts Part III: Fe Oxide-Containing Melts

Science.gov (United States)

Thibodeau, Eric; Gheribi, Aimen E.; Jung, In-Ho

2016-04-01

As part III of this series, the model is extended to iron oxide-containing melts. All available experimental data in the FeO-Fe2O3-Na2O-K2O-MgO-CaO-MnO-Al2O3-SiO2 system were critically evaluated based on the experimental condition. The variations of FeO and Fe2O3 in the melts were taken into account by using FactSage to calculate the Fe2+/Fe3+ distribution. The molar volume model with unary and binary model parameters can be used to predict the molar volume of the molten oxide of the Li2O-Na2O-K2O-MgO-CaO-MnO-PbO-FeO-Fe2O3-Al2O3-SiO2 system in the entire range of compositions, temperatures, and oxygen partial pressures from Fe saturation to 1 atm pressure.

Energy Transfer between U(VI) and Eu(III) Ions Adsorbed on a Silica Surface

Energy Technology Data Exchange (ETDEWEB)

Park, K. K.; Cha, W.; Cho, H. R.; Im, H. J.; Jung, E. C.; Song, K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Understanding of chemical behavior of actinide in a groundwater flow is important for assessing the possibility of their migration with water flows in a radioactive waste disposal site. Uranium is ubiquitous in the environment and a major actinide in a nuclear fuel cycle. Americium and curium having isotopes of long half life are minor actinides in a spent fuel. If a minor actinide coexists with uranium in a groundwater flow, some interactions between them could be expected such as minor actinide adsorption onto uranium precipitates and competition with each other for an adsorption to a mineral surface site. Eu(III) ion is frequently used as a chemical analogue of Am(III) and Cm(III) ions in a migration chemistry. The luminescent spectra of U(VI) and Eu(III) ions show a dependency on the coordination symmetry around them, and the changes in intensity or bandwidth of spectra can yield valuable information on their local environment. The luminescent lifetime also strongly depends on the coordination environment, and its measurement is valuable in probe studies on micro-heterogeneous systems. The excited U(VI) ion can be quenched through Stern.Volmer process, hydrolysis of excited species, exciplex formation, electron transfer or energy transfer. In case of U(VI)-Eu(III) system, the interaction between two ions can be studied by measuring the effect of Eu(III) ion on the quenching of U(VI) ion luminescence. There are only a few investigations on the interaction between an excited U(VI) ion and a lanthanide(III) ion. In perchlorate solution, the energy transfer to Eu(III) ion occurred only in solutions of pH>3.87. In this study, the quenching of U(VI) luminescence by Eu(III) on a silica surface was measured. The results will be discussed on the basis of a chemical interaction between them

A review on the use of gas and steam turbine combined cycles as prime movers for large ships. Part I: Background and design

International Nuclear Information System (INIS)

Haglind, Fredrik

2008-01-01

The aim of the present paper is to review the prospects of using combined cycles as prime movers for large ships, like, container ships, tankers and bulk carriers. The paper is divided into three parts of which this paper constitutes Part I. Here, the environmental and human health concerns of international shipping are outlined. The regulatory framework relevant for shipping and the design of combined cycles are discussed. In Part II, previous work and experience are reviewed, and an overview of the implications of introducing combined cycles as prime movers is included. In Part III, marine fuels are discussed and the pollutant emissions of gas turbines are compared with those of two-stroke, slow-speed diesel engines. Environmental effects of shipping include contributions to the formation of ground-level ozone, acidification, eutrophication and climate impact. Tightening environmental regulations limit the fuel sulphur content and pollutant emissions. For moderate live steam pressures, a vertical HRSG of drum-type mounted directly over the gas turbine, is suggested to be a viable configuration that minimizes ground floor and space requirements

Fuel salt and container material studies for MOSART transforming system

Energy Technology Data Exchange (ETDEWEB)

Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Institute, Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Institute of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A. [Institute of Technical Physics, Snezhinsk (Russian Federation)

2013-07-01

A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

Post irradiation test report of irradiated DUPIC simulated fuel

International Nuclear Information System (INIS)

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

2001-12-01

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

Evaluation of biodiesel fuel and a diesel oxidation catalyst in an underground metal mine : Part 3 : Biological and chemical characterization

Energy Technology Data Exchange (ETDEWEB)

Bagley, S.T. [Michigan Technological Univ., Houghton, MI (United States). Dept. of Biological Sciences; Gratz, L.D. [Michigan Technological Univ., Houghton, MI (United States). Dept. of Mechanical Engineering-Engineering Mechanics

1998-07-24

A collaborative, international, multidisciplinary effort led to the evaluation of the effects of using a 50 per cent biodiesel fuel blend and an advanced-type diesel oxidation catalyst (DOC) on underground metal mine air quality. The location selected for the field trials was the Creighton Mine 3 in Sudbury, Ontario, operated by Inco. Specifically, part 3 of the study evaluated the effects of using a biodiesel blend fuel on potentially health-related diesel particulate matter (DPM) components, with a special emphasis on polynuclear aromatic hydrocarbons (PAH), nitro-PAH, and mutagenic activity. High volume sampler filters containing submicrometer particles were examined, and comparisons made for DPM and DPM component concentrations. The downwind concentrations of DPM were reduced by 20 per cent with the use of the blend biodiesel fuel as compared with the number 2 diesel fuel with an advanced-type DOC. Significant reductions in solids (up to 30 per cent) and up to 75 per cent in the case of mutagenic activity were noted. Significant reductions in the DPM components potentially harmful to human health should result from the use of this blended fuel combined with an advanced-type DOC in an underground environment. 23 refs., 19 tabs.

Automotive Fuel Processor Development and Demonstration with Fuel Cell Systems

Energy Technology Data Exchange (ETDEWEB)

Nuvera Fuel Cells

2005-04-15

processor subsystems (fuel reformer, CO cleanup, and exhaust cleanup) that were small enough to integrate on a vehicle and (2) evaluating the fuel processor system performance for hydrogen production, efficiency, thermal integration, startup, durability and ability to integrate with fuel cells. Nuvera carried out a three-part development program that created multi-fuel (gasoline, ethanol, natural gas) fuel processing systems and investigated integration of fuel cell / fuel processor systems. The targets for the various stages of development were initially based on the goals of the DOE's Partnership for New Generation Vehicles (PNGV) initiative and later on the Freedom Car goals. The three parts are summarized below with the names based on the topic numbers from the original Solicitation for Financial Assistance Award (SFAA).

Extended fuel cycle length

International Nuclear Information System (INIS)

Bruyere, M.; Vallee, A.; Collette, C.

1986-09-01

Extended fuel cycle length and burnup are currently offered by Framatome and Fragema in order to satisfy the needs of the utilities in terms of fuel cycle cost and of overall systems cost optimization. We intend to point out the consequences of an increased fuel cycle length and burnup on reactor safety, in order to determine whether the bounding safety analyses presented in the Safety Analysis Report are applicable and to evaluate the effect on plant licensing. This paper presents the results of this examination. The first part indicates the consequences of increased fuel cycle length and burnup on the nuclear data used in the bounding accident analyses. In the second part of this paper, the required safety reanalyses are presented and the impact on the safety margins of different fuel management strategies is examined. In addition, systems modifications which can be required are indicated

Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward

International Nuclear Information System (INIS)

Turchi, P.E.; Kaufman, L.; Fluss, M.J.

2008-01-01

The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermochemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenge are not insurmountable and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER

Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward

Energy Technology Data Exchange (ETDEWEB)

Turchi, P E; Kaufman, L; Fluss, M J

2008-11-10

The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermochemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenge are not insurmountable and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

Part-load performance and emissions of a spark ignition engine fueled with RON95 and RON97 gasoline: Technical viewpoint on Malaysia’s fuel price debate

International Nuclear Information System (INIS)

Mohamad, Taib Iskandar; How, Heoy Geok

2014-01-01

Highlights: • Recent Malaysia’s gasoline price hike affects mass perception and vehicle sales. • Effects of RON95 and RON97 on a representative engine was experimentally studied. • RON95 produced better torque, power, fuel efficiency and lower NO x . • RON97 gasoline resulted in lower BSFC and lower emissions of CO 2 , CO and HC. • Performance-emission-price cross-analysis indicated RON95 as the better option. - Abstract: Due to world crude oil price hike in the recent years, many countries have experienced increase in gasoline price. In Malaysia, where gasoline are sold in two grades; RON95 and RON97, and fuel price are regulated by the government, gasoline price have been gradually increased since 2009. Price rise for RON97 is more significant. By 2014, its per liter price is 38% more than that of RON95. This has resulted in escalated dissatisfaction among the mass. People argued they were denied from using a better fuel (RON97). In order to evaluate the claim, there is a need to investigate engine response to these two gasoline grades. The effect of gasoline RON95 and RON97 on performance and exhaust emissions in spark ignition engine was investigated on a representative engine: 1.6L, 4-cylinder Mitsubishi 4G92 engine with CR 11:1. The engine was run at constant speed between 1500 and 3500 rpm with 500 rpm increment at various part-load conditions. The original engine ECU, a hydraulic dynamometer and control, a combustion analyzer and an exhaust gas analyzer were used to determine engine performance, cylinder pressure and emissions. Results showed that RON95 produced higher engine performance for all part-load conditions within the speed range. RON95 produced on average 4.4% higher brake torque, brake power, brake mean effective pressure as compared to RON97. The difference in engine performance was more significant at higher engine speed and loads. Cylinder pressure and ROHR were evaluated and correlated with engine output. With RON95, the engine

Fuel vapor pressure (FVAPRS)

International Nuclear Information System (INIS)

Mason, R.E.

1979-04-01

A subcode (FVAPRS) is described which calculates fuel vapor pressure. This subcode was developed as part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The fuel vapor pressure subcode (FVAPRS), is presented and a discussion of literature data, steady state and transient fuel vapor pressure equations and estimates of the standard error of estimate to be expected with the FVAPRS subcode are included

Storage device for a long nuclear reactor fuel element and/or a long nuclear reactor fuel element part

International Nuclear Information System (INIS)

Vogt, M.; Schoenwitz, H.P.; Dassbach, W.

1986-01-01

The storage device can be erected in a dry storage room for new fuel elements and also in a storage pond for irradiated fuel elements. It consists of shells, which are arranged vertically and which have a lid. A suspension for the fuel element is provided on the underside of the lid, which acts as a support against squashing or bending in case of vertical forces acting (earthquake). (DG) [de

Separation of Am(III) from Eu(III) by mixtures of triazynylbipyridine and bis(dicarbollide) extractants. The composition of the metal complexes extracted

International Nuclear Information System (INIS)

Narbutt, J.; Krejzler, J.

2006-01-01

Separation of trivalent actinides, in particular americium and curium, from lanthanides is an important step in an advanced partitioning process for future reprocessing of spent nuclear fuels. The use of soft donor (N and S) ligands makes it possible to separate the two groups of elements, probably because of the more covalent character in the complexes with actinides compared to the lanthanides. The aim of present work was to study solvent extraction of Am(III) and Eu(III) in a similar system with diethylhemi-BTP and COSAN: protonated bis(chlorodicarbollido)cobalt(III) or commo-3,3-cobalta-bis(8,9,12-trichlora-1,2-dicarbaclosododecaborane)ic acid. The present research was focused on both the determination of conditions for the separation of 241 Am(III) from 152 Eu in aqueous nitrate solution by using a synergistic extraction system and on the modelling of the process by slope analysis. Obtained values of the separation factors supported by the computer modelling permitted drawing the conclusions on the mechanism of the process and on the structure of extracted species

Oxidation of dibenzothiophene as a model substrate for the removal of organic sulphur from fossil fuels by iron(III ions generated from pyrite by Acidithiobacillus ferrooxidans

Directory of Open Access Journals (Sweden)

VLADIMIR P. BESKOSKI

2007-06-01

Full Text Available Within this paper a new idea for the removal of organically bonded sulphur from fossil fuels is discussed. Dibenzothiophene (DBT was used as a model compound of organicmolecules containing sulphur. This form of (biodesulphurization was performed by an indirect mechanism in which iron(III ions generated from pyrite by Acidithiobacillus ferrooxidans performed the abiotic oxidation. The obtained reaction products, dibenzothiopene sulfoxide and dibenzothiophene sulfone, are more soluble in water than the basic substrate and the obtained results confirmed the basic hypothesis and give the posibility of continuing the experiments related to application of this (biodesulphurization process.

Fabrication of PWR fuel assembly and CANDU fuel bundle

International Nuclear Information System (INIS)

Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

1980-01-01

For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

Nuclear Fuel Reprocessing

International Nuclear Information System (INIS)

Simpson, Michael F.; Law, Jack D.

2010-01-01

This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

Cost evaluation of a commercial-scale DUPIC fuel fabrication facility (Part I) -Summary

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-08-01

A conceptual design of a commercial scale DUPIC fuel fabrication facility was initiated to provide some insights into the costs associated with construction, operation, and decommissioning. The primary conclusion of this report is that it is feasible to design, license, construct, test, and operate a facility that will process 400 MTHE/yr of spent PWR fuel and reconfigure the fuel into CANDU fuel bundles at a reasonable unit cost of the fuel material. Although DUPIC fuel fabrication by vibropacking method is clearly cheaper than that of the pellet method, the feasibility of vibropac technology for DUPIC fuel fabrication and use of vibroac fuel in CANDU reactors may has to be studied in depth in order to use as an alternative to the conventional pellet fuel method. Especially, there are some questions on meeting the CANDU requirements in thermal and mechanical terms as well as density of fuel. Wherever possible, this report used representative costs of currently available technologies as the bases for cost estimation. It should also be noted that the conceptual design and cost information contained in this report was extracted from the public domain and general open literature. Later studies have to focus on other important areas of concern such as safety, security, safeguards, process optimization etc. 7 figs., 6 tabs. (Author)

Oral Assessment Kit, Levels II & III. Draft.

Science.gov (United States)

Agrelo-Gonzalez, Maria; And Others

The assessment packet includes a series of oral tests to help develop speaking as an integral part of second language instruction at levels II and III. It contains: 8 mini-tests for use at level II; 9 mini-tests for use at level III; a rating scale and score sheet masters for evaluating performance on these tests; and a collection of suggested…

Introductory remarks about the international fuel cycle

International Nuclear Information System (INIS)

Wolfe, B.

1989-01-01

The reason why nuclear power has promise is because of the promise of its fuel cycle. The fuel cycle is in fairly good shape and has demonstrated the characteristics of good economics, good general characterization, and good maintenance of the various parts of the fuel cycle. The thermal recycling of fuel is an area in which the economics have changed to the point that, at least in many parts of the world, it's no longer economical

Fabrication of fuel elements on the basis of increased concentration fuel composition

International Nuclear Information System (INIS)

Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

2004-01-01

As a part of Russian Program RERTR Reduced Enrichment for Research and Test Reactors), at NCCP, Inc. jointly with the State Scientific Centre VNIINM the mastering in industrial environment of design and fabrication process of fuel elements (FE) with increased concentration fuel compositions is performed. Fuel elements with fuel composition on the basis of dioxide uranium with nearly 4 g/cm 3 fuel concentration have been produced thus confirming the principal possibility of fuel enrichment reduction down to 20% for research reactors which were built up according to the projects of the former USSR, by increasing the oxide fuel concentration in fuel assemblies (FAs). The form and geometrical dimensions of FEs and FAs shall remain unchanged, only uranium mass in FA shall be increased. (author)

Nondestructive nuclear measurement in the fuel cycle. Part 1

International Nuclear Information System (INIS)

Lyoussi, A.

2005-01-01

Nondestructive measurement techniques are today widely used in practically all steps of the fuel cycle. This article is devoted to the presentation of the control and characterization needs and to the main passive nondestructive nuclear methods used: 1 - nondestructive nuclear measurement, needs and motivation: nuclear fuel cycle, nondestructive nuclear measurements (passive and active methods), comments; 2 - main passive nondestructive nuclear measurement methods: gamma spectroscopy (principle, detectors, electronic systems, data acquisition and signal processing, domains of application, main limitations), passive neutronic measurements (needs and motivations, neutron detectors, total neutronic counting, neutronic coincidences counting, neutronic multiplicities counting, comments). (J.S.)

Enhancing BWR proliferation resistance fuel with minor actinides

Science.gov (United States)

Chang, Gray S.

2009-03-01

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in

Alternative Practices to Improve Surface Fleet Fuel Efficiency

Science.gov (United States)

2014-09-01

through changes in procedures and operational modifications. iENCON uses BBLs/hr (barrels per hour) to evaluate the change in fuel efficiency (Pehlivan...policies and procedures that can be changed to continue the Navy’s efforts in the reduction of fuel consumption. Chapter III addresses drift...and four main engines. In a “full power” lineup all four engines are online. In a “split plant” lineup two engines remain online, one per shaft

14 CFR Appendix M to Part 25 - Fuel Tank System Flammability Reduction Means

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel Tank System Flammability Reduction... 25—Fuel Tank System Flammability Reduction Means M25.1Fuel tank flammability exposure requirements. (a) The Fleet Average Flammability Exposure of each fuel tank, as determined in accordance with...

Technical realisation of the VISA-3 project, Parts I-II, Part I; Tehnicka realizacija projekta VISA-3, I-II deo, I Deo

Energy Technology Data Exchange (ETDEWEB)

Pavicevic, M; Smokovic, Z [Institute of Nuclear Sciences Boris Kidric, Odeljenje za reaktorsku eksperimentalnu tehniku, Vinca, Beograd (Serbia and Montenegro)

1966-11-15

This task is related to irradiation of reactor materials (steel, Al, MgO, Al{sub 2}O{sub 3}, ets.) at higher temperatures (200-500 deg C) in the fast neutron flux. These conditions would be more realistic to real reactor conditions than the conditions achieved within VISA-2 project. The experimental space will be the same as in VISA-2 project, i.e. refurbished reactor channels and within the fuel elements. The irradiation capsule will be leak tight with thermal isolation layer and supplied with electric heater to enable temperature variation.

21st Century Renewable Fuels, Energy, and Materials

Energy Technology Data Exchange (ETDEWEB)

Berry, K. Joel [Kettering Univ., Flint, MI (United States); Das, Susanta K. [Kettering Univ., Flint, MI (United States)

2012-11-29

The objectives of this project were multi-fold: (i) conduct fundamental studies to develop a new class of high temperature PEM fuel cell material capable of conducting protons at elevated temperature (180°C), (ii) develop and fabricate a 5k We novel catalytic flat plate steam reforming process for extracting hydrogen from multi-fuels and integrate with high-temperature PEM fuel cell systems, (iii) research and develop improved oxygen permeable membranes for high power density lithium air battery with simple control systems and reduced cost, (iv) research on high energy yield agriculture bio-crop (Miscanthus) suitable for reformate fuel/alternative fuel with minimum impact on human food chain and develop a cost analysis and production model, and (v) develop math and science alternative energy educator program to include bio-energy and power.

Taxing carbon in fuels

International Nuclear Information System (INIS)

Arnold, Rob

2000-01-01

It is argued that both the Climate Change Levy and the fuel duty tax are outdated even before they are implemented. Apparently, the real problems are not in the bringing of road fuels into the scope of the Climate Change Levy but in introducing reforms to improve integration of greenhouse gases and taxation. Both fuel duty and the Levy are aimed at maximising efficiency and reducing air pollution. The system as it stands does not take into account the development of a market where the management and trading of carbon and greenhouse gases may jeopardise the competitiveness of UK businesses. It is argued that an overhaul of climate and emissions-related law is necessary. The paper is presented under the sub-headings of (i) a fixation on energy; (ii) no focus on CO 2 ; (iii) carbon markets - beyond the levy and (iv) tax structure. (UK)

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

Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor; Normativa aplicable y desarrollo de experimentos de vigilancia de aproximacion a criticidad en el reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J L; Aguilar H, F; Rivero G, T; Sainz M, E [Instituto nacional de Investigaciones Nucleares, Departamento de Automatizacion, A.P. 18-1027, Col. Escandon, 11801 Mexico D.F. (Mexico)

2000-07-01

In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

Thermal sensation and comfort models for non-uniform and transient environments, part III: Whole-body sensation and comfort

Energy Technology Data Exchange (ETDEWEB)

Zhang, Hui; Arens, Edward; Huizenga, Charlie [Center for the Built Environment, UC Berkeley (United States); Han, Taeyoung [General Motors Company (United States)

2010-02-15

A three-part series presents the development of models for predicting the local thermal sensation (Part I) and local thermal comfort (Part II) of different parts of the human body, and also the whole-body sensation and comfort (Part III) that result from combinations of local sensation and comfort. The models apply to sedentary activities in a range of environments: uniform and non-uniform, stable and transient. They are based on diverse findings from the literature and from body-part-specific human subject tests in a climate chamber. They were validated against a test of automobile passengers. The series is intended to present the models' rationale, structure, and coefficients, so that others can test them and develop them further as additional empirical data becomes available. A) The whole-body (overall) sensation model has two forms, depending on whether all of the body's segments have sensations effectively in the same direction (e.g warm or cool), or whether some segments have sensations opposite to those of the rest of the body. For each, individual body parts have different weights for warm versus cool sensations, and strong local sensations dominate the overall sensation. If all sensations are near neutral, the overall sensation is close to the average of all body sensations. B) The overall comfort model also has two forms. Under stable conditions, people evaluate their overall comfort by a complaint-driven process, meaning that when two body parts are strongly uncomfortable, no matter how comfortable the other body parts might be, the overall comfort will be near the discomfort level of the two most uncomfortable parts. When the environmental conditions are transient, or people have control over their environments, overall comfort is better than that of the two most uncomfortable body parts. This can be accounted for by adding the most comfortable vote to the two most uncomfortable ones. (author)

Development of LWR fuel performance code FEMAXI-6

International Nuclear Information System (INIS)

Suzuki, Motoe

2006-01-01

LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

Fuel Handbook[Wood and other renewable fuels

Energy Technology Data Exchange (ETDEWEB)

Stroemberg, Birgitta [TPS Termiska Processer AB, Nykoeping (SE)] (ed.)

2006-03-15

This handbook on renewable fuels is intended for power and heat producers in Sweden. This fuel handbook provides, from a plant owner's perspective, a method to evaluate different fuels on the market. The fuel handbook concerns renewable fuels (but does not include household waste) that are available on the Swedish market today or fuels that have potential to be available within the next ten years. The handbook covers 26 different fuels. Analysis data, special properties, operating experiences and literature references are outlined for each fuel. [Special properties, operating experiences and literature references are not included in this English version] The handbook also contains: A proposed methodology for introduction of new fuels. A recommendation of analyses and tests to perform in order to reduce the risk of problems is presented. [The recommendation of analyses and tests is not included in the English version] A summary of relevant laws and taxes for energy production, with references to relevant documentation. [Only laws and taxes regarding EU are included] Theory and background to evaluate a fuel with respect to combustion, ash and corrosion properties and methods that can be used for such evaluations. Summary of standards, databases and handbooks on biomass fuels and other solid fuels, and links to web sites where further information about the fuels can be found. The appendices includes: A methodology for trial firing of fuels. Calculations procedures for, amongst others, heating value, flue gas composition, key number and free fall velocity [Free fall velocity is not included in the English version]. In addition, conversion routines between different units for a number of different applications are provided. Fuel analyses are presented in the appendix. (The report is a translation of parts of the report VARMEFORSK--911 published in 2005)

Fuel safety research 2000

Energy Technology Data Exchange (ETDEWEB)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

In April 1999, the Fuel Safety Research Laboratory was newly established as a part of reorganization of the Nuclear Safety Research Center, JAERI. The new laboratory was organized by combining three pre-existing laboratories, Reactivity Accident Laboratory, Fuel Reliability Laboratory, and a part of Severe Accident Research Laboratory. The Fuel Safety Research Laboratory becomes to be in charge of all fuel safety research in JAERI. Various experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of following five research groups corresponding to each research fields; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). The research activities in year 2000 produced many important data and information. They are, for example, failure of high burnup BWR fuel rod under RIA conditions, data on the behavior of hydrided Zircaloy cladding under LOCA conditions and FP release data from VEGA experiments at very high temperature/pressure condition. This report summarizes the outline of research activities and major outcomes of the research executed in 2000 in the Fuel Safety Research Laboratory. (author)

Guidebook on spent fuel storage

International Nuclear Information System (INIS)

1984-01-01

The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland

ASME codification of ductile cast iron cask for transport and storage of spent nuclear fuel

International Nuclear Information System (INIS)

Saegusa, Toshiari; Arai, Taku

2012-01-01

The CRIEPI has been executing research and development on ductile cast iron cask for transport and storage of spent nuclear fuel in order to diversify options of the casks. Based on the research results, the CRIEPI proposed materials standards (Section II) and structural design standards (Section III) for the ductile cast iron cask to the authoritative and international ASME (American Society of Mechanical Engineers) Codes. For the Section II, the CRIEPI proposed the JIS G 5504 material with additional requirement prohibiting repair of cast body by welding, etc. as well as the ASTM A874 material to the Part A. In addition, the CRIEPI proposed design stress allowables, physical properties (thermal conductivity, modulus of elasticity, etc.), and external pressure chart to the Part D. For the Section III, the CRIEPI proposed a fracture toughness requirement of the ductile cast iron cask at -40degC to WB and WC of Division 3. Additionally, the CRIEPI proposed a design fatigue curve of the ductile cast iron cask to Appendix of Division 1. This report describes the outline of the proposed standards, their bases, and the deliberation process in order to promote proper usage of the code, future improvement, etc. (author)

Determination of free acid in U(VI)-Al(III) solution by Gran plot titration

International Nuclear Information System (INIS)

Suh, Moo Yul; Lee, Chang Heon; Sohn, Se Chul; Kim, Jung Suk; Kim, Won Ho; Eom, Tae Yoon

1999-01-01

The determination method of free acid in spent U-Al nuclear fuel solutions by Gran plot titration was described. Effect of U(VI) and Al(III) on the alkalimetric titration of nitric acid was investigation in oxalate complexing media as well as in noncomplexing media. Positive biases were observed in both titration media when the end-point was estimated by the Gran plot method. It was found that the cause of the bias was U(VI) in the oxalate complexing media, but Al(III) in the noncomplexing media. The relative error was less than 1% in the titration of 0.1 M HNO 3 at a U(VI):Al(III):H + mole ratio of up to 2:12:1 as long as the pH of the oxalate titration media was sustained to be below 5.0 at the beginning of titration. The method was successfully applied to the determination of nitric acid in a solution of HANARO reactor fuel with U:Al mole ratio of 1:6

Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

International Nuclear Information System (INIS)

Black, J.E.; King, F.D.

1979-01-01

Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

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)

CANDU fuel quality and how it is achieved

International Nuclear Information System (INIS)

Gacesa, M.; Quarrington, G.R.; Tarasuk, W.R.; Carrick, I.R.; Pawliw, J.; McGregor, G.; Debnam, H.R.; Proos, L.

1980-07-01

In this three part presentation CANDU fuel quality is reviewed from the point of view of a designer/operator and a fabricator. In Part 'A' fuel performance and quality considerations are discussed from the point of view of a designer-operator. In Parts 'B' and 'C' fuel quality is reviewed from the point of view of a fabricator. The presentation was divided in this way to convey the 'team effort' attitude which exists in the Canadian program; the team effort which is an essential part of the CANDU story. (auth)

Spent Fuel Management Newsletter. No. 1

International Nuclear Information System (INIS)

1990-03-01

This Newsletter has been prepared in accordance with the recommendations of the International Regular Advisory Group on Spent Fuel Management and the Agency's programme (GC XXXII/837, Table 76, item 14). The main purpose of the Newsletter is to provide Member States with new information about the state-of-the-art in one of the most important parts of the nuclear fuel cycle - Spent Fuel Management. The contents of this publication consists of two parts: (1) IAEA Secretariat contribution -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, etc. (2) Country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage, treatment of spent fuel, some aspects of uranium and plutonium recycling, etc. The IAEA expects to publish the Newsletter once every two years between the publications of the Regular Advisory Group on Spent Fuel Management. Figs and tabs

The Nuclear Fuel Cycle

International Nuclear Information System (INIS)

2011-08-01

This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

Cermet fuel reactors

International Nuclear Information System (INIS)

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.

1987-01-01

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

Fuel consumption from vehicles of China until 2030 in energy scenarios

International Nuclear Information System (INIS)

Zhang Qingyu; Tian Weili; Zheng Yingyue; Zhang Lili

2010-01-01

Estimation of fuel (gasoline and diesel) consumption for vehicles in China under different long-term energy policy scenarios is presented here. The fuel economy of different vehicle types is subject to variation of government regulations; hence the fuel consumption of passenger cars (PCs), light trucks (Lts), heavy trucks (Hts), buses and motor cycles (MCs) are calculated with respect to (i) the number of vehicles, (ii) distance traveled, and (iii) fuel economy. On the other hand, the consumption rate of alternative energy sources (i.e. ethanol, methanol, biomass-diesel and CNG) is not evaluated here. The number of vehicles is evaluated using the economic elastic coefficient method, relating to per capita gross domestic product (GDP) from 1997 to 2007. The Long-range Energy Alternatives Planning (LEAP) system software is employed to develop a simple model to project fuel consumption in China until 2030 under these scenarios. Three energy consumption decrease scenarios are designed to estimate the reduction of fuel consumption: (i) 'business as usual' (BAU); (ii) 'advanced fuel economy' (AFE); and (iii) 'alternative energy replacement' (AER). It is shown that fuel consumption is predicted to reach 992.28 Mtoe (million tons oil equivalent) with the BAU scenario by 2030. In the AFE and AER scenarios, fuel consumption is predicted to be 734.68 and 600.36 Mtoe, respectively, by 2030. In the AER scenario, fuel consumption in 2030 will be reduced by 391.92 (39.50%) and 134.29 (18.28%) Mtoe in comparison to the BAU and AFE scenarios, respectively. In conclusion, our models indicate that the energy conservation policies introduced by governmental institutions are potentially viable, as long as they are effectively implemented.

State-of-the-Art Report on Multi-scale Modelling of Nuclear Fuels

International Nuclear Information System (INIS)

Bartel, T.J.; Dingreville, R.; Littlewood, D.; Tikare, V.; Bertolus, M.; Blanc, V.; Bouineau, V.; Carlot, G.; Desgranges, C.; Dorado, B.; Dumas, J.C.; Freyss, M.; Garcia, P.; Gatt, J.M.; Gueneau, C.; Julien, J.; Maillard, S.; Martin, G.; Masson, R.; Michel, B.; Piron, J.P.; Sabathier, C.; Skorek, R.; Toffolon, C.; Valot, C.; Van Brutzel, L.; Besmann, Theodore M.; Chernatynskiy, A.; Clarno, K.; Gorti, S.B.; Radhakrishnan, B.; Devanathan, R.; Dumont, M.; Maugis, P.; El-Azab, A.; Iglesias, F.C.; Lewis, B.J.; Krack, M.; Yun, Y.; Kurata, M.; Kurosaki, K.; Largenton, R.; Lebensohn, R.A.; Malerba, L.; Oh, J.Y.; Phillpot, S.R.; Tulenko, J. S.; Rachid, J.; Stan, M.; Sundman, B.; Tonks, M.R.; Williamson, R.; Van Uffelen, P.; Welland, M.J.; Valot, Carole; Stan, Marius; Massara, Simone; Tarsi, Reka

2015-10-01

Fuels is to document the development of multi-scale modelling approaches for fuels in support of current fuel optimisation programmes and innovative fuel designs. The objectives of the effort are: - assess international multi-scale modelling approaches devoted to nuclear fuels from the atomic to the macroscopic scale in order to share and promote such approaches; - address all types of fuels: both current (mainly oxide fuels) and advanced fuels (such as minor actinide containing oxide, carbide, nitride, or metal fuels); - address key engineering issues associated with each type of fuel; - assess the quality of existing links between the various scales and list needs for strengthening multi-scale modelling approaches; - identify the most relevant experimental data or experimental characterisation techniques that are missing for validation of fuel multi-scale modelling; - promote exchange between the actors involved at various scales; - promote exchange between multi-scale modelling experts and experimentalists; - exchange information with other expert groups of the WPMM. This report is organised as follows: - Part I lays out the different classes of phenomena relevant to nuclear fuel behaviour. Each chapter is further divided into topics relevant for each class of phenomena. - Part II is devoted to a description of the techniques used to obtain material properties necessary for describing the phenomena and their assessment. - Part III covers details relative to the principles and limits behind each modelling/computational technique as a reference for more detailed information. Included within the appropriate sections are critical analyses of the mid- and long-term challenges for the future (i.e., approximations, methods, scales, key experimental data, characterisation techniques missing or to be strengthened)

Spectroscopic investigations on the complexation of Cm(III) and Eu(III) with organic model ligands and their binding mode in human urine (in vitro)

International Nuclear Information System (INIS)

Heller, Anne

2011-01-01

In case of incorporation, trivalent actinides (An(III)) and lanthanides (Ln(III)) pose a serious health risk to humans. An(III) are artificial, highly radioactive elements which are mainly produced during the nuclear fuel cycle in nuclear power plants. Via hazardous accidents or nonprofessional storage of radioactive waste, they can be released in the environment and enter the human food chain. In contrast, Ln(III) are nonradioactive, naturally occurring elements with multiple applications in technique and medicine. Consequently it is possible that humans get in contact and incorporate both, An(III) and Ln(III). Therefore, it is of particular importance to elucidate the behaviour of these elements in the human body. While macroscopic processes such as distribution, accumulation and excretion are studied quite well, knowledge about the chemical binding form (speciation) of An(III) and Ln(III) in various body fluids is still sparse. In the present work, for the first time, the speciation of Cm(III) and Eu(III) in natural human urine (in vitro) has been investigated spectroscopically and the formed complex identified. For this purpose, also basic investigations on the complex formation of Cm(III) and Eu(III) in synthetic model urine as well as with the urinary relevant, organic model ligands urea, alanine, phenylalanine, threonine and citrate have been performed and the previously unknown complex stability constants determined. Finally, all experimental results were compared to literature data and predictions calculated by thermodynamic modelling. Since both, Cm(III) and Eu(III), exhibit unique luminescence properties, particularly the suitability of time-resolved laser-induced fluorescence spectroscopy (TRLFS) could be demonstrated as a method to investigate these metal ions in untreated, complex biofluids. The results of this work provide new scientific findings on the biochemical reactions of An(III) and Ln(III) in human body fluids on a molecular scale and

Spacer for supporting fuel element boxes

International Nuclear Information System (INIS)

Wild, E.

1979-01-01

A spacer plate unit arranged externally on each side and at a predetermined level of a polygonal fuel element box for mutually supporting, with respect to one another, a plurality of the fuel element boxes forming a fuel element bundle, is formed of a first and a second spacer plate part each having the same length and the same width and being constituted of unlike first and second materials, respectively. The first and second spacer plate parts of the several spacer plate units situated at the predetermined level are arranged in an alternating continuous series when viewed in the peripheral direction of the fuel element box, so that any two spacer plate units belonging to face-to-face oriented sides of two adjoining fuel element boxes in the fuel element bundle define interfaces of unlike materials

Environmental systems analysis of biogas systems-Part I: Fuel-cycle emissions

International Nuclear Information System (INIS)

Boerjesson, Pal; Berglund, Maria

2006-01-01

Fuel-cycle emissions of carbon dioxide (CO 2 ), carbon oxide (CO), nitrogen oxides (NO x ), sulphur dioxide (SO 2 ), hydrocarbons (HC), methane (CH 4 ), and particles are analysed from a life-cycle perspective for different biogas systems based on six different raw materials. The gas is produced in large- or farm-scale biogas plants, and is used in boilers for heat production, in turbines for co-generation of heat and electricity, or as a transportation fuel in light- and heavy-duty vehicles. The analyses refer mainly to Swedish conditions. The levels of fuel-cycle emissions vary greatly among the biogas systems studied, and are significantly affected by the properties of the raw material digested, the energy efficiency of the biogas production, and the status of the end-use technology. For example, fuel-cycle emission may vary by a factor of 3-4, and for certain gases by up to a factor of 11, between two biogas systems that provide an equivalent energy service. Extensive handling of raw materials, e.g. ley cropping or collection of waste-products such as municipal organic waste, is often a significant source of emissions. Emission from the production phase of the biogas exceeds the end-use emissions for several biogas systems and for specific emissions. Uncontrolled losses of methane, e.g. leakages from stored digestates or from biogas upgrading, increase the fuel-cycle emissions of methane considerably. Thus, it is necessary to clearly specify the biogas production system and end-use technology being studied in order to be able to produce reliable and accurate data on fuel-cycle emission

Hydrophilic 2,9-bis-triazolyl-1,10-phenanthroline ligands enable selective Am(iii) separation: a step further towards sustainable nuclear energy.

Science.gov (United States)

Edwards, Alyn C; Mocilac, Pavle; Geist, Andreas; Harwood, Laurence M; Sharrad, Clint A; Burton, Neil A; Whitehead, Roger C; Denecke, Melissa A

2017-05-02

The first hydrophilic, 1,10-phenanthroline derived ligands consisting of only C, H, O and N atoms for the selective extraction of Am(iii) from spent nuclear fuel are reported herein. One of these 2,9-bis-triazolyl-1,10-phenanthroline (BTrzPhen) ligands combined with a non-selective extracting agent, was found to exhibit process-suitable selectivity for Am(iii) over Eu(iii) and Cm(iii), providing a clear step forward.

National report of the Slovak Republic compiled in terms of the Join on the safety of spent fuel management and on the safety of radwaste management

International Nuclear Information System (INIS)

Jurina, V.; Viktory, D.; Petrik, T.; Sovcik, J.; Suess, J.; Tomek, J.; Lukacovic, J.; Ivan, J.; Ziakova, M.; Metke, E.; Pospisil, M.; Turner, M.; Homola, J.; Vaclav, J.; Bystricka, S.; Barbaric, M.; Horvath, J.; Betak, J.; Mihaly, B.; Adamovsky, V.; Baloghova, A.; Orihel, M.; Vasina, D.; Balaz, J.; Misovicova, D.; Vrtoch, M.; Mlcuch, J.; Granak, P.; Meleg, J.; Bardy, M.; Gogoliak, J.

2011-08-01

The National safety report of the Slovak Republic on the safety of spent fuel management and on the safety of radwaste management in 2011 is presented. These activities in the safety of spent fuel management and radioactive waste management in the Slovak Republic are reported under the headings: (A) Introduction; B) Concept for spent nuclear fuel management (SNF) and radwaste management (RAW); (C) Scope of application of the convention; (D) Spent fuel management and radioactive waste (RAW) management facilities; (E) Legislation and regulation; (F) General safety provisions; (G) Safety of spent fuel management; (H) Safety of radioactive waste (RAW) management; (I) Transboundary movement of spent nuclear fuel and radioactive waste; (J) Disused sealed sources; (K) Planned measures to improve safety; (L) Communication with the public; (M) Annexes. Annexes consists of following parts: I. List of nuclear facilities for spent fuel and RAW management. II. Limits of radioactive material discharges into atmosphere and hydrosphere. III. List of nuclear installations in decommissioning. IV. Inventory of stored spent nuclear fuel. V. Inventory of stored RAW. VI. List of national laws, decrees and guidelines. VII. List of international expert reports (including safety reports). VIII. List of authors.

Spectroscopic studies on the interaction of europium(III) and curium(III) with components of the human mucosa

Energy Technology Data Exchange (ETDEWEB)

Wilke, Claudia; Barkleit, Astrid [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Chemistry of the F-Elements

2016-07-01

To evaluate the health risks of lanthanides (Ln) and radiotoxic actinides (An) in case of ingestion accidents etc., investigations into the chemical reactions of these metals in the human gastrointestinal tract are necessary. Our previous study revealed that mucin, an important part of the protective mucosa layer in the digestive system, shows a strong interaction with Eu(III). Based on these results, the present study focuses on the components of this glycoprotein and identified N-acetylneuraminic acid (NANA) as the dominant binding carbohydrate of mucin. TRLFS measurements suggest the formation of a 1: 1 complex with log β of 3.2 ± 0.1 for Eu(III) and 3.3 ± 0.1 for Cm(III), respectively.

An economic analysis of transportation fuel policies in Brazil: Fuel choice, land use, and environmental impacts

International Nuclear Information System (INIS)

Nuñez, Hector M.; Önal, Hayri

2016-01-01

Brazil uses taxes, subsidies, and blending mandates as policy instruments to manage and stabilize its transportation fuel markets. The fuel sector has been very dynamic in recent years due to frequent policy adjustments and variable market conditions. In this paper, we use a price endogenous economic simulation model to analyze the impacts of such policy adjustments under various challenging conditions in the global ethanol and sugar markets. Our analysis specifically focuses on Brazilian producers' supply responses, consumers' driving demand and fuel choice, ethanol trade, land use, greenhouse gas emissions, and social welfare. The model results show that (i) under a low ethanol blending rate, conventional vehicles would be driven significantly less while flex-fuel and ethanol-dedicated vehicles would not be affected significantly; (ii) lowering the fuel taxes adversely affects the competitiveness of sugarcane ethanol against gasoline blends, thus lowering producers' surplus; and (iii) while a reduction in fuel taxes is advantageous in terms of overall social welfare, it has serious environmental impacts by increasing the GHG emissions from transportation fuels consumed in Brazil. - Highlights: • We examine the economic and environmental impacts of Brazilian fuel policies. • We also analyze impacts under different sugar and ethanol markets conditions. • Lowering blending rate reduces distance driven by conventional cars. • Lowering fuel tax rates affects competitiveness of ethanol against gasoline blend. • Reducing fuel tax rates has dramatic environmental impacts by increasing emissions.

Irradiated uranium reprocessing, Final report - I-IV, Part III

International Nuclear Information System (INIS)

Gal, I.

1961-12-01

This third part of the final report include the following: Annex 5 - device for opening the cover; Annex 6 - inner part of the device for sampling of the radioactive solution; Annex 7 - outer part of the device for sampling of the radioactive solution; Annex 8 - pneumatic taps [sr

HTGR Fuel performance basis

International Nuclear Information System (INIS)

Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

1982-05-01

The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Science.gov (United States)

2010-01-01

... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION... Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities 1. Public health... facilities for the temporary storage of highlevel radioactive wastes, may be located on privately owned...

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.

Mixing ratio sensor for alcohol mixed fuel

Energy Technology Data Exchange (ETDEWEB)

Miyata, Shigeru; Matsubara, Yoshihiro

1987-08-24

In order to improve the combustion efficiency of an internal combustion engine using gasoline-alcohol mixed fuel and to reduce harmful substance in its exhaust gas, it is necessary to control strictly the air-fuel ratio to be supplied and the ignition timing. In order to detect the mixing ratio of the mixed fuel, a mixing ratio sensor has so far been proposed to detect the above mixing ratio by casting a ray of light to the mixed fuel and utilizing a change of critical angle associated with the change of the composition of the fluid of the mixed fuel. However, because of the arrangement of its transparent substance in the fuel passage with the sealing material in between, this sensor invited the leakage of the fluid due to deterioration of the sealing material, etc. and its cost became high because of too many parts to be assembled. In view of the above, in order to reduce the number of parts, to lower the cost of parts and the assembling cost and to secure no fluid leakage from the fuel passage, this invention formed the above fuel passage and the above transparent substance both concerning the above mixing ratio sensor in an integrated manner using light transmitting resin. (3 figs)

Isothiocyanato complexes of Gd(III), Tb(III), Dy(III) and Ho(III) with 2-(2'-pyridyl)benzimidazole

Energy Technology Data Exchange (ETDEWEB)

Mishra, A; Singh, V K

1982-01-01

Six-coordinated complexes of the type (Ln(PyBzH)/sub 2/NCS.H/sub 2/O) (NCS)/sub 2/.nH/sub 2/O/mC/sub 2/H/sub 5/OH (Ln = Gd(III), Tb(III), Dy(III) and Ho(III), n=1-2; m=1) have been prepared from Ln(NCS)/sub 6//sup 3 -/. The room temperature magnetic moment values confirm the terpositive state of the lanthanide ions. Infrared spectra suggest the N-coordination of thiocyanate group. Electronic spectral studies of Tb(III), Dy(III) and Ho(III) complexes have been made in terms of LSJ term energies. 13 refs.

Removable fuel assembly for nuclear reactor

International Nuclear Information System (INIS)

Dubief, J.M.; Bonnamour, M.

1984-01-01

To facilitate the replacement of one or more fuel rods, taking into account the fact the operations are remote operations and under several meters of water, the following invention is presented. The fuel assembly is composed of a bundle of canned fuel pencils maintened on a structure which includes ends linked by spacer tubes. These tubes are fixed to one end in such a manner they are removable. For this, the plug of each tube has a plane stop surface on the end part and a conic coupling and guiding plug cooperating with a truncated bearing of the end part. Flat parts made on the cone allow to stop the tube rotating [fr

Structural analysis of fuel handling systems

Energy Technology Data Exchange (ETDEWEB)

Lee, L S.S. [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1997-12-31

The purpose of this paper has three aspects: (i) to review `why` and `what` types of structural analysis, testing and report are required for the fuel handling systems according to the codes, or needed for design of a product, (ii) to review the input requirements for analysis and the analysis procedures, and (iii) to improve the communication between the analysis and other elements of the product cycle. The required or needed types of analysis and report may be categorized into three major groups: (i) Certified Stress Reports for design by analysis, (ii) Design Reports not required for certification and registration, but are still required by codes, and (iii) Design Calculations required by codes or needed for design. Input requirements for structural analysis include: design, code classification, loadings, and jurisdictionary boundary. Examples of structural analysis for the fueling machine head and support structure are given. For improving communication between the structural analysis and the other elements of the product cycle, some areas in the specification of design requirements and load rating are discussed. (author). 6 refs., 1 tab., 4 figs.

Structural analysis of fuel handling systems

International Nuclear Information System (INIS)

Lee, L.S.S.

1996-01-01

The purpose of this paper has three aspects: (i) to review 'why' and 'what' types of structural analysis, testing and report are required for the fuel handling systems according to the codes, or needed for design of a product, (ii) to review the input requirements for analysis and the analysis procedures, and (iii) to improve the communication between the analysis and other elements of the product cycle. The required or needed types of analysis and report may be categorized into three major groups: (i) Certified Stress Reports for design by analysis, (ii) Design Reports not required for certification and registration, but are still required by codes, and (iii) Design Calculations required by codes or needed for design. Input requirements for structural analysis include: design, code classification, loadings, and jurisdictionary boundary. Examples of structural analysis for the fueling machine head and support structure are given. For improving communication between the structural analysis and the other elements of the product cycle, some areas in the specification of design requirements and load rating are discussed. (author). 6 refs., 1 tab., 4 figs

Automated Fuel Element Closure Welding System

International Nuclear Information System (INIS)

Wahlquist, D.R.

1993-01-01

The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

Nuclear Fuel Cycle Information System. A directory of nuclear fuel cycle facilities. 2009 ed

International Nuclear Information System (INIS)

2009-04-01

The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities, published online as part of the Integrated Nuclear Fuel Cycle Information System (iNFCIS: http://www-nfcis.iaea.org/). This is the fourth hardcopy publication in almost 30 years and it represents a snapshot of the NFCIS database as of the end of 2008. Together with the attached CD-ROM, it provides information on 650 civilian nuclear fuel cycle facilities in 53 countries, thus helping to improve the transparency of global nuclear fuel cycle activities

Proceedings of the Annual Meeting of the Association for Education in Journalism and Mass Communication (83rd, Phoenix, Arizona, August 9-12, 2000). Miscellaneous, Part III.

Science.gov (United States)

Association for Education in Journalism and Mass Communication.

The Miscellaneous, part III section of the proceedings contains the following 11 papers: "The Relationship between Health and Fitness Magazine Reading and Eating-Disordered Weight-Loss Methods among High School Girls" (Steven R. Thomsen, Michelle M. Weber, and Lora Beth Brown); "A Practical Exercise for Teaching Ethical Decision…

Burnable poison calculations for Mk.III gas-cooled reactor systems

Energy Technology Data Exchange (ETDEWEB)

Gubbins, M E

1971-02-15

A method of calculating the reactivity and burn-up hisotry of a Mk.III GCR system containing burnable poisons has been described. The method allows for poison-fuel interaction. Using the method it has been shown that burn-up of the poison under a constant incident flux can give errors of the order of 1-2 niles. A calculation using the method described will take about 50% longer than a straightforward fuel burn-up calculation in the same number of groups. The multi-cell approach has a potential for handling greater geometrical complexity. It is intended to compare the method against experiment as soon as suitable experimental results become available.

Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method

International Nuclear Information System (INIS)

2004-01-01

This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 2. of ISO 7097 describes procedures for determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with cerium(IV) and ISO 7097-1 uses a titration with potassium dichromate

Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method

International Nuclear Information System (INIS)

2004-01-01

This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 1. of ISO 7097 describes procedures for the determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with potassium dichromate and ISO 7097-2 uses a titration with cerium(IV)

New camera systems for fuel services

International Nuclear Information System (INIS)

Hummel, W.; Beck, H.J.

2010-01-01

AREVA NP Fuel Services have many years of experience in visual examination and measurements on fuel assemblies and associated core components by using state of the art cameras and measuring technologies. The used techniques allow the surface and dimensional characterization of materials and shapes by visual examination. New enhanced and sophisticated technologies for fuel services f. e. are two shielded color camera systems for use under water and close inspection of a fuel assembly. Nowadays the market requirements for detecting and characterization of small defects (lower than the 10th of one mm) or cracks and analyzing surface appearances on an irradiated fuel rod cladding or fuel assembly structure parts have increased. Therefore it is common practice to use movie cameras with higher resolution. The radiation resistance of high resolution CCD cameras is in general very low and it is not possible to use them unshielded close to a fuel assembly. By extending the camera with a mirror system and shielding around the sensitive parts, the movie camera can be utilized for fuel assembly inspection. AREVA NP Fuel Services is now equipped with such kind of movie cameras. (orig.)

National report of the Slovak Republic compiled in terms of the join convention on the safety of spent fuel management and on the safety of radwaste management

International Nuclear Information System (INIS)

Jurina, V.; Viktory, D.; Petrik, T.; Sovcik, J.; Suess, J.; Tomek, J.; Lukacovic, J.; Ivan, J.; Ziakova, M.; Metke, E.; Pospisil, M.; Turner, M.; Homola, J.; Vaclav, J.; Bystricka, S.; Barbaric, M.; Horvath, J.; Betak, J.; Mihaly, B.; Adamovsky, V.; Baloghova, A.; Orihel, M.; Vasina, D.; Balaz, J.; Misovicova, D.; Vrtoch, M.; Mlcuch, J.; Granak, P.; Meleg, J.; Bardy, M.; Gogoliak, J.

2011-08-01

The National safety report of the Slovak Republic on the safety of spent fuel management and on the safety of radwaste management in 2011 is presented. These activities in the safety of spent fuel management and radioactive waste management in the Slovak Republic are reported under the headings: (A) Introduction; B) Concept for spent nuclear fuel management (SNF) and radwaste management (RAW); (C) Scope of application of the convention; (D) Spent fuel management and radioactive waste (RAW) management facilities; (E) Legislation and regulation; (F) General safety provisions; (G) Safety of spent fuel management; (H) Safety of radioactive waste (RAW) management; (I) Transboundary movement of spent nuclear fuel and radioactive waste; (J) Disused sealed sources; (K) Planned measures to improve safety; (L) Communication with the public; (M) Annexes. Annexes consists of following parts: I. List of nuclear facilities for spent fuel and RAW management. II. Limits of radioactive material discharges into atmosphere and hydrosphere. III. List of nuclear installations in decommissioning. IV. Inventory of stored spent nuclear fuel. V. Inventory of stored RAW. VI. List of national laws, decrees and guidelines. VII. List of international expert reports (including safety reports). VIII. List of authors.

Optimisation of polypyrrole/Nafion composite membranes for direct methanol fuel cells

International Nuclear Information System (INIS)

Zhu Jun; Sattler, Rita R.; Garsuch, Arnd; Yepez, Omar; Pickup, Peter G.

2006-01-01

Acidic and neutral Nafion[reg] 115 perfluorosulphonate membranes have been modified by in situ polymerization of pyrrole using Fe(III) and H 2 O 2 as oxidizing agents, in order to decrease methanol crossover in direct methanol fuel cells. Improved selectivities for proton over methanol transport and improved fuel cell performances were only obtained with membranes that were modified while in the acid form. Use of Fe(III) as the oxidizing agent can produce a large decrease in methanol crossover, but causes polypyrrole deposition on the surface of the membrane. This increases the resistance of the membrane, and leads to poor fuel cell performances due to poor bonding with the electrodes. Surface polypyrrole deposition can be minimized, and surface polypyrrole can be removed, by using H 2 O 2 . The use of Nafion in its tetrabutylammonium form leads to very low methanol permeabilities, and appears to offer potential for manipulating the location of polypyrrole within the Nafion structure

Nuclear fuel assembly

International Nuclear Information System (INIS)

Betten, P.R.

1976-01-01

Under the invention the fuel assembly is particularly suitable for liquid metal cooled fast neutron breeder reactors. Hence, according to the invention a fuel assembly cladding includes inward corrugations with respect to the remainder of the cladding according to a recurring pattern determined by the pitch of the metal wire helically wound round the fuel rods of the assembly. The parts of the cladding pressed inwards correspond to the areas in which the wire encircling the peripheral fuel rods is generally located apart from the cladding, thereby reducing the play between the cladding and the peripheral fuel rods situated in these areas. The reduction in the play in turn improves the coolant flow in the internal secondary channels of the fuel assembly to the detriment of the flow in the peripheral secondary channels and thereby establishes a better coolant fluid temperature profile [fr

Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor

International Nuclear Information System (INIS)

Gonzalez M, J.L.; Aguilar H, F.; Rivero G, T.; Sainz M, E.

2000-01-01

In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

International Nuclear Information System (INIS)

Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

2007-01-01

The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

Development of nuclear fuel cycle technologies

International Nuclear Information System (INIS)

Suzuoki, Akira; Matsumoto, Takashi; Suzuki, Kazumichi; Kawamura, Fumio

1995-01-01

In the long term plan for atomic energy that the Atomic Energy Commission decided the other day, the necessity of the technical development for establishing full scale fuel cycle for future was emphasized. Hitachi Ltd. has engaged in technical development and facility construction in the fields of uranium enrichment, MOX fuel fabrication, spent fuel reprocessing and so on. In uranium enrichment, it took part in the development of centrifuge process centering around Power Reactor and Nuclear Fuel Development Corporation (PNC), and took its share in the construction of the Rokkasho uranium enrichment plant of Japan Nuclear Fuel Service Co., Ltd. Also it cooperates with Laser Enrichment Technology Research Association. In Mox fuel fabrication, it took part in the construction of the facilities for Monju plutonium fuel production of PNC, for pellet production, fabrication and assembling processes. In spent fuel reprocessing, it cooperated with the technical development of maintenance and repair of Tokai reprocessing plant of PNC, and the construction of spent fuel stores in Rokkasho reprocessing plant is advanced. The centrifuge process and the atomic laser process of uranium enrichment are explained. The high reliability of spent fuel reprocessing plants and the advancement of spent fuel reprocessing process are reported. Hitachi Ltd. Intends to exert efforts for the technical development to establish nuclear fuel cycle which increases the importance hereafter. (K.I.)

Application of contact mechanics for fretting damage of fuel rod: part 1 influence functions and numerical method

International Nuclear Information System (INIS)

Kim, H. K.; Yoon, K. H.; Kang, H. S.; Song, G. N.

1998-01-01

For the analysis of the fretting problem of the fuel rods, present paper(Part I) shows the numerical method developed for evaluating the stresses on the contact surfaces between the fuel rods and the spacer grids. Theory of Contact Mechanics was incorporated. Contact area was regarded as a plane strain condition, so plane problem was taken into consideration. Normal stress profile on the contact surface was assumed to be Hertzian. As for the direction of the shear load, a closed load path, e.g. load increase in transverse increase in axial decrease in transverse decrease in axial increase in transverse increase in axial direction was considered for simulating the rod vibration in a reactor core. Partial slip problem was consulted. As for the numerical method, a triangular traction element was utilized and the corresponding influence functions were evaluated. Numerical program has been implemented for the present analysis, of which the validity was verified by comparing the Mindlin-Cattaneo solution

Potential sites for a spent unreprocessed fuel facility (SURFF), southwesten part of the Nevada Test Site

International Nuclear Information System (INIS)

Hoover, D.L.; Eckel, E.B.; Ohl, J.P.

1978-01-01

In the absence of specific criteria, the topography, geomorphology, and geology of Jackass Flats and vicinity in the southwestern part of the Nevada Test Site are evaluated by arbitrary guidelines for a Spent Unreprocessed Fuel Facility. The guidelines include requirements for surface slopes of less than 5%, 61 m of alluvium beneath the site, an area free of active erosion or deposition, lack of faults, a minimum area of 5 km 2 , no potential for flooding, and as many logistical support facilities as possible. The geology of the Jackass Flats area is similar to the rest of the Nevada Test Site in topographic relief (305-1,200 m), stratigraphy (complexly folded and faulted Paleozoic sediments overlain by Tertiary ash-flow tuffs and lavas overlain in turn by younger alluvium), and structure (Paleozoic thrust faults and folds, strike-slip faults, proximity to volcanic centers, and Basin and Range normal faults). Of the stratigraphic units at the potential Spent Unreprocessed Fuel Facility site in Jackass Flats, only the thickness and stability of the alluvium are of immediate importance. Basin and Range faults and a possible extension of the Mine Mountain fault need further investigation. The combination of a slope map and a simplified geologic and physiographic map into one map shows several potential sites for a Spent Unreprocessed Fuel Facility in Jackass Flats. The potential areas have slopes of less than 5% and contain only desert pavement or segmented pavement--the two physiographic categories having the greatest geomorphic and hydraulic stability. Before further work can be done, specific criteria for a Spent Unreprocessed Fuel Facility site must be defined. Following criteria definition, potential sites will require detailed topographic and geologic studies, subsurface investigations (including geophysical methods, trenching, and perhaps shallow drilling for faults in alluvium), detailed surface hydrologic studies, and possibly subsurface hydrologic studies

A tri-generation system based on polymer electrolyte fuel cell and desiccant wheel – Part A: Fuel cell system modelling and partial load analysis

International Nuclear Information System (INIS)

Najafi, Behzad; De Antonellis, Stefano; Intini, Manuel; Zago, Matteo; Rinaldi, Fabio; Casalegno, Andrea

2015-01-01

Highlights: • A mathematical model for a PEMFC based cogeneration system is developed. • Developed model is validated using the available experimental data. • Performance of the plant at full load conditions is investigated. • Performance indices while applying two different modifications are determined. • System’s performance with and without modifications at partial loads is investigated. - Abstract: Polymer Electrolyte Membrane Fuel Cell (PEMFC) based systems have recently received increasing attention as a viable alternative for meeting the residential electrical and thermal demands. However, as the intermittent demand profiles of a building can only be addressed by a tri-generative unit which can operate at partial loads, the variation of performance of the system at partial loads might affect its corresponding potential benefits significantly. Nonetheless, no previous study has been carried out on assessing the performance of this type of tri-generative systems in such conditions. The present paper is the first of a two part study dedicated to the investigation of the performance of a tri-generative system in which a PEMFC based system is coupled with a desiccant wheel unit. This study is focused on evaluating the performance of the PEMFC subsystem while operating at partial loads. Accordingly, a detailed mathematical model of the fuel cell subsystem is first developed and validated using the experimental data obtained from the plant’s and the fuel cell stack’s manufacturer. Next, in order to increase the performance of the plant, two modifications have been proposed and the resulting performance at partial load have been determined. The obtained results demonstrate that applying both modifications results in increasing the electrical efficiency of the plant by 5.5%. It is also shown that, while operating at partial loads, the electrical efficiency of the plant does not significantly change; the fact which corresponds to the trade-off between

Improvement of fuel injection system of locomotive diesel engine.

Science.gov (United States)

Li, Minghai; Cui, Hongjiang; Wang, Juan; Guan, Ying

2009-01-01

The traditional locomotive diesels are usually designed for the performance of rated condition and much fuel will be consumed. A new plunger piston matching parts of fuel injection pump and injector nozzle matching parts were designed. The experimental results of fuel injection pump test and diesel engine show that the fuel consumption rate can be decreased a lot in the most of the working conditions. The forced lubrication is adopted for the new injector nozzle matching parts, which can reduce failure rate and increase service life. The design has been patented by Chinese State Patent Office.

Enhancing VVER annular proliferation resistance fuel with minor actinides

International Nuclear Information System (INIS)

Chang, G. S.

2007-01-01

Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 2 38Pu and 2 40Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 2 37Np and 2 41Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 2 38Pu /Pu. For future advanced nuclear systems, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors

Safety of nuclear fuel cycle facilities. Safety requirements

International Nuclear Information System (INIS)

2008-01-01

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

Update of the Used Fuel Dispositon Campaign Implementation Plan

Energy Technology Data Exchange (ETDEWEB)

McMahon, Kevin A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mackinnon, Robert James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Swift, Peter N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Birkholzer, Jens T. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2014-10-01

This Update to the Used Fuel Disposition Campaign Implementation Plan provides summary level detail describing how the Used Fuel Disposition Campaign (UFDC) supports achievement of the overarching mission and objectives of the Department of Energy Office of Nuclear Energy Fuel Cycle Technologies Program, building on work completed in this area since 2009. This implementation plan begins with the assumption of target dates that are set out in the January 2013 DOE Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (http://energy.gov/downloads/strategy-management-and-disposal-used-nuclearfuel- and-high-level-radioactive-waste). These target dates and goals are summarized in section III. This implementation plan will be maintained as a living document and will be updated as needed in response to available funding and progress in the Used Fuel Disposition Campaign and the Fuel Cycle Technologies Program.

Techno-economic assessment of fuel cell vehicles for India

International Nuclear Information System (INIS)

Manish S; Rangan Banerjee

2006-01-01

This paper compares four alternative vehicle technologies for a typical small family car in India (Maruti 800) - two conventional i) Petrol driven internal combustion (IC) engine, ii) Compressed natural gas (CNG) driven IC engine and two based on proton exchange membrane (PEM) fuel cells with different storage iii) Compressed hydrogen storage and iv) Metal hydride (FeTi) storage. Each technology option is simulated in MATLAB using a backward facing algorithm to calculate the force and power requirement for the Indian urban drive cycle. The storage for the CNG and the fuel cell vehicles is designed to have driving range of 50% of the existing petrol vehicle. The simulation considers the part load efficiency vs. load characteristics for the computed ratings of the IC engine and the fuel cell. The analysis includes the transmission efficiency, motor efficiency and storage efficiencies. The comparison criteria used are the primary energy consumption (MJ/km), the cost (Rs./km) obtained by computing the annualized life cycle cost and dividing this by the annual vehicle travel and carbon dioxide emissions (g/km). For the primary energy analysis the energy required for extraction, processing of the fuel is also included. For the fuel cell vehicles, it is assumed that hydrogen is produced from natural gas through steam methane reforming. It is found that the fuel cell vehicles have the lowest primary energy consumption (1.3 MJ/km) as compared to the petrol and CNG vehicles (2.3 and 2.5 MJ/km respectively). The cost analysis is done based on existing prices in India and reveals that the CNG vehicle has the lowest cost (2.3 Rs./km) as compared to petrol (4.5 Rs./km). The fuel cell vehicles have a higher cost of 26 Rs./km mainly due to the higher fuel cell system cost (93% of the total cost). The CO 2 emissions are lowest for the fuel cell vehicle with compressed hydrogen storage (98 g/km) as compared to the petrol vehicle (162 g/km). If the incremental annual cost of the fuel

Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Ding, Ming, E-mail: dingm2005@gmail.com [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, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

2014-02-15

Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

CERCA'S experience in UMO fuel manufacturing

International Nuclear Information System (INIS)

Jarousse, Ch.; Lavastre, Y.; Grasse, M.

2003-01-01

Considered as a suitable solution for non-proliferation and reprocessing purposes, UMo fuel has been chosen and studied by the RERTR program since 1996. Involved in the RERTR fuel developments since 1978, with more than 20 years of U 3 SI 2 fuel production, and closely linked to the French Commissariat a l'Energie Atomique, CERCA was able to define properly, from the beginning, the right R and D actions plan for UMo fuel development. CERCA has already demonstrated during the last 4 years its ability to manufacture plates and fuel elements with high density UMo fuel. UMo full size plates produced for 4 irradiation experiments in 3 European reactors afforded us a unique experience. In addition, as a main part of our R and D effort, we have always studied in depth a key part of the CERCA process outline which is the plate rolling stage. After some preliminary investigation in order to define the phenomenological model describing the behavior of the fuel core when rolling, we have developed a rolling digital simulator. (author)

Fuel assemblies

International Nuclear Information System (INIS)

Echigoya, Hironori; Nomata, Terumitsu.

1983-01-01

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

Materials for low-temperature fuel cells

CERN Document Server

Ladewig, Bradley; Yan, Yushan; Lu, Max

2014-01-01

There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in Low-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in low-temperature fuel cells. A related book will cover key materials in high-temperature fuel cells. The two books form part

Complexes of 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III)

International Nuclear Information System (INIS)

Ferenc, W.; Bernat, M; Gluchowska, H.W.; Sarzynski, J.

2010-01-01

The complexes of 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III) have been synthesized as polycrystalline hydrated solids, and characterized by elemental analysis, spectroscopy, magnetic studies and also by X-ray diffraction and thermogravimetric measurements. The analysed complexes have the following colours: violet for Nd(III), white for Gd(III) and cream for Ho(III) compounds. The carboxylate groups bind as bidentate chelating (Ho) or bridging ligands (Nd, Gd). On heating to 1173K in air the complexes decompose in several steps. At first, they dehydrate in one step to form anhydrous salts, that next decompose to the oxides of respective metals. The gaseous products of their thermal decomposition in nitrogen were also determined and the magnetic susceptibilities were measured over the temperature range of 76-303K and the magnetic moments were calculated. The results show that 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III) are high-spin complexes with weak ligand fields. The solubility value in water at 293K for analysed 4-chlorophenoxyacetates is in the order of 10 -4 mol/dm 3 . (author)

Dental compensation for skeletal Class III malocclusion by isolated extraction of mandibular teeth. Part 1: Occlusal situation 12 years after completion of active treatment.

Science.gov (United States)

Zimmer, Bernd; Schenk-Kazan, Sarah

2015-05-01

The purpose of this work was to statistically evaluate the outcomes achieved by isolated extraction of mandibular teeth (second premolars or first molars) for Class III compensation. Part A of the study dealt with the quality of outcomes at the end of active treatment, using weighted Peer Assessment Rating (PAR) scores determined on the basis of casts for 25 (14 female and 11 male) consecutive patients aged 16 ± 1.7 years at the time of debonding. These results were compared to the scores in a randomly selected control group of 25 (14 female and 11 male) patients who were 14.7 ± 1.9 years old at debonding. Part B evaluated the long-term stability of the outcomes based on 12 (all of them female) patients available for examination after a mean of 11.8 years. The mean weighted PAR scores obtained in both study parts were analyzed for statistical differences using a two-tailed paired Student's t-test at a significance level of p ≤ 0.05. Mean weighted PAR scores of 4.76 ± 3.94 and 3.92 ± 3.44 were obtained in the Class III extraction group and the control group, respectively, at the end of active treatment. This difference was not significant (p = 0.49). Among the 12 longitudinal patients, the mean score increased from 4 ± 3.46 at debonding to 6.25 ± 3.67 by the end of the 11.8-year follow-up period. This difference was significant (p = 0.0008). Treatment of Class III anomalies by isolated extraction of lower premolars or molars can yield PAR scores similar to those achieved by standard therapies. These scores, while increasing significantly, remained at a clinically acceptable level over 11.8 years. Hence this treatment modality--intended for cases that border on requiring orthognathic surgery--may also be recommended from a long-term point of view.

Fuel taxes and road expenditures: making the link

International Nuclear Information System (INIS)

Derkson, W.W.; Shurvell, S.J.

1999-11-01

This document reports on a study undertaken at the request of the United Grain Growers regarding government fuel tax revenue and the relationship to expenditures on roads. The account of fuel tax revenues was compiled from data collected from several different sources, as was the case for the road expenditures at the federal, provincial, local and modal levels. The emphasis was placed on the effects of fuel taxes on grain handling and transportation in the Prairie provinces. The authors presented fuel tax revenues broken down by mode of transportation and by province. The document was divided as follows: the first part was the introduction with the second part dealing with fuel tax rates and policies. In the third part, the topic of fuel tax and road related revenues was examined. Part four discussed road expenditures. The authors concluded that Transport Canada has traditionally represented the most important federal link to provincial highway infrastructure. It was noted that 4.2 billion dollars in road fuel taxes were collected by the federal government in 1998/1999, and of that amount, 198 million dollars, or 4.7 per cent, was reinvested in the National Highway System in programs managed by Transport Canada. Nearly one dollar on roads is spent by provincial governments on a Prairie-wide basis for every dollar collected in road fuel taxes, with Alberta spending the most and Saskatchewan spending the least. 15 tabs

Assessment of fuel concepts

International Nuclear Information System (INIS)

Bailey, W.J.; Barner, J.O.

1978-01-01

The relative merits of various LWR UO 2 fuel concepts with the potential for improved power-ramping capability were qualitatively assessed. In the evaluation, it was determined that of the various concepts being considered, those that presently possess an adequately developed experience base include annular pellets, cladding coated with graphite on the inner surface, and packed-particle fuel. Therefore, these were selected for initial evaluation as part of the Fuel Performance Improvement Program. For this program, graphite-coated cladding is being used in conjunction with annular pellet fuel as one of the concepts with the anticipation of gaining the advantage of the combined improvements. The report discusses the following: the criteria used to evaluate the candidate fuel concepts; a comparison of the concepts selected for irradiation with the criteria, including a general description of their experience bases; and a general discussion of other candidate concepts, including identifying those which may be considered for out-of-reactor evaluation as part of this program, those for which the results of other programs will be monitored, and those which have been deleted from further consideration at this time

Fuel damage during off-normal transients in metal-fueled fast reactors

International Nuclear Information System (INIS)

Kramer, J.M.; Bauer, T.H.

1990-01-01

Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor; Diseno y construccion del SIPPING para combustibles del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

2003-07-01

The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

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...

Fuel cycle based safeguards

International Nuclear Information System (INIS)

De Montmollin, J.M.; Higinbotham, W.A.; Gupta, D.

1985-07-01

In NPT safeguards the same model approach and absolute-quantity inspection goals are applied at present to all similar facilities, irrespective of the State's fuel cycle. There is a continuing interest and activity on the part of the IAEA in new NPT safeguards approaches that more directly address a State's nuclear activities as a whole. This fuel cycle based safeguards system is expected to a) provide a statement of findings for the entire State rather than only for individual facilities; b) allocate inspection efforts so as to reflect more realistically the different categories of nuclear materials in the different parts of the fuel cycle and c) provide more timely and better coordinated information on the inputs, outputs and inventories of nuclear materials in a State. (orig./RF) [de

Nuclear fuel performance in boiling water reactors

International Nuclear Information System (INIS)

Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

1981-01-01

A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

A competitive thorium fuel cycle for pressurized water reactors of current technology

International Nuclear Information System (INIS)

Galperin, A.; Radkowsky, A.; Todosow, M.

2002-01-01

Two important issues may influence the development and public acceptance of the nuclear power worldwide: a reduction of proliferation potential and spent fuel disposal requirements of the nuclear fuel cycle. Both problems may be addressed effectively by replacement of uranium by thorium fertile part of the fuel. A practical and competitive fuel design to satisfy the described design objectives and constraints may be achieved by seed-blanket core, proposed by A. Radkowsky and implemented in Shippingport reactors. The main idea is to separate spatially the uranium part of the core (seed) from the thorium part of the core (blanket), and thus allow two separate fuel management routes for uranium and thorium parts of the fuel. The uranium part (seed) is optimized to supply neutrons to the subcritical thorium blanket. The blanket is designed to generate and bum insitu 233 U. (author)

49 CFR 393.65 - All fuel systems.

Science.gov (United States)

2010-10-01

... motor vehicles or for the operation of auxiliary equipment installed on, or used in connection with, motor vehicles. (b) Location. Each fuel system must be located on the motor vehicle so that— (1) No part... will not contact any part of the exhaust or electrical systems of the vehicle, except the fuel level...

Participation in the IAEA Coordinated Research Project Fumex III: Final Report of AREVA NP

International Nuclear Information System (INIS)

2013-01-01

After the Coordinated Research Project (CRP) FUMEXII, participants asked for a new exercise within an IAEA CRP. This CRP started in December 2008 in Vienna with the first Research Coordination Meeting (RCM). The CRP is titled ''Improvement of Computer Codes Used for Fuel Behaviour Simulation FUMEX III''. The object of FUMEX III were the improvement of fuel rod performance codes for modeling high burnup phenomena in modern fuel. This includes transient behavior, as well as mechanical interaction between pellet and cladding and, in progression to the FUMEX II exercise, fission gas release during various conditions (steady state, load follow, transient). AREVA NP agreed on participating in this exercise under the IAEA research agreement no. 15369 and expressed interest in the modeling of pelletclad mechanical interactions as well as fission gas release under steady state and transient conditions. In this exercise AREVA NP used its new global fuel rod code GALILEO, which is still under development (formerly known under the project name COPERNIC 3). During a Consultants Meeting potential topics and a proposed selection of cases have been prepared, which were discussed during the 1st Research Coordination Meeting (RCM) in Vienna in December 2008. During the discussions a number of additional cases motivated by the participants have been identified. Finally, a case table has been agreed upon, which included several cases for the different topics. Most of the cases have been based on the International Fuel Performance Experiments (IFPE) database, but additional cases have been provided during the exercise (e.g., the AREVA idealized case

Transportation of spent nuclear fuels

International Nuclear Information System (INIS)

Meguro, Toshiichi

1976-01-01

The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)

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.

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

Science.gov (United States)

2010-08-01

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

Winter fuels report

Energy Technology Data Exchange (ETDEWEB)

1990-11-29

The Winter Fuels Report is intended to provide concise, timely information to the industry, the press, policymakers, consumers, analysts, and state and local governments on the following topics: distillate fuel oil net production, imports and stocks for all PADD's and product supplied on a US level; propane net product supplied on a US level; propane net production, imports and stocks for Petroleum Administration for Defense Districts (PADD) I, II, and III; natural gas supply and disposition and underground storage for the United States and consumption for all PADD's; residential and wholesale pricing data for propane and heating oil for those states participating in the joint Energy Information Administration (EIA)/State Heating Oil and Propane Program; crude oil and petroleum price comparisons for the United States and selected cities; and US total heating degree-days by city. 27 figs, 12 tabs.

Fuel formula for lighters

Energy Technology Data Exchange (ETDEWEB)

Iwayama, I.; Iwayama, A.

1982-04-10

A fuel formula that includes a homogenous mixture of benzine, aromatic ether oils, perfume and other perfuming agents, as well as the lowest possible aliphatic alcohol as a component solvent, surfactant, and possibly, a soluble pigment that colors the formula an appropriate color. This formula is used as an aromatic fuel for cigarette lights. The ether oils can be musk, amber, camomille, lavender, mint, anise, rose, camphor, and other aromatic oils; the perfuming agents are: geraniol, linalool, menthol, camphor, benzyl or phenetyl alcohols, phenylacetaldehyde, vanillin, coumarin, and so forth; the pigments are: beta-carotene, sudan dyes, etc.; the low aliphatic alcohols are EtOH, iso-PrOH. Example: 70 parts benzine, 10 parts EtOH, 15 parts oxide mezithylene and 5 parts borneol form a clear liquid that has a camphor aroma when it is lit.

Nalco Fuel Tech

Energy Technology Data Exchange (ETDEWEB)

Michalak, S.

1995-12-31

The Nalco Fuel Tech with its seat at Naperville (near Chicago), Illinois, is an engineering company working in the field of technology and equipment for environmental protection. A major portion of NALCO products constitute chemical materials and additives used in environmental protection technologies (waste-water treatment plants, water treatment, fuel modifiers, etc.). Basing in part on the experience, laboratories and RD potential of the mother company, the Nalco Fuel Tech Company developed and implemented in the power industry a series of technologies aimed at the reduction of environment-polluting products of fuel combustion. The engineering solution of Nalco Fuel Tech belong to a new generation of environmental protection techniques developed in the USA. They consist in actions focused on the sources of pollutants, i.e., in upgrading the combustion chambers of power engineering plants, e.g., boilers or communal and/or industrial waste combustion units. The Nalco Fuel Tech development and research group cooperates with leading US investigation and research institutes.

Thermodynamic analysis of advanced fuels for fast breeder reactors

International Nuclear Information System (INIS)

Srivastava, D.; Garg, S.P.; Goswami, G.L.

1990-01-01

Six phase fields of interest in the M-C-N system (M= mixed U/Pu) with oxygen as impurity are i) U 1-x3 Pu x3 (=M)+ U 1-x1 Pu x1 C 1-y-z N y O z (= MCN O), ii)C+ U 1 x2 Pu x2 Csub(1.5) (=MCsub(1.5)), iii) MCsub(1.5) + MCNO, iv) C+MCNO, v) UN (1.5) + MCNO and vi) C + UNsub(1.5) + MCNO. In the present work a detailed thermodynamic analysis has been carried out for all the six phase fields existing in the system with x 1 , 1-y-z and y are varying from 0.0 to 1.0 and z as impurity from 0.0 to 0.15 at temperature between 1500K to 2000K. In the first part, composition of the phases in the different phase fields have been calculated as a function of overall composition of the fuel and temperature. In the second part, thermodynamic properties such as partial pressures of N 2 (g), O 2 (g), CO(g), Pu(g), U(g), PuO(g), UO(g), UC 2 (g) and PuC 2 (g) species and carbon potential of the fuel have been calculated as a function of compositions x 1 , y and z at different temperatures. Results obtained are discus sed in detail and compared with the reported measured data. Hitherto, thermodynamic properties for all the phase fields of M-C-N-O system have not been reported. (a uthor). 54 tabs., 13 figs., 24 refs

IAEA programme on nuclear fuel cycle and materials technologies - 2009

International Nuclear Information System (INIS)

Killeen, J.

2009-01-01

In this paper a brief description and the main objectives of IAEA Programme B on Nuclear fuel cycle are given. The following Coordinated Research Projects: 1) Delayed Hydride Cracking (DHC); 2) Structural Materials Radiation Effects (SMoRE); 3) Water Chemistry (FUWAC) and 4) Fuel Modelling (FUMEX-III) are shortly described. The data collected by the IAEA Expert Group of Fuel Failures in Water Cooled Reactors including information about fuel assembly damage that did not result in breach of the fuel rod cladding, such as assembly bow or crud deposition an the experience with these unexpected fuel issues shows that they can seriously affect plant operations, and it is clear that concerns about reliability in this area are of similar importance today as fuel rod failures, at least for LWR fuel are discussed. Detection, examination and analysis of fuel failures and description of failures and mitigation measures as well as preparation of a Monograph on Zirconium including an overview of Zirconium for nuclear applications, including extraction, forming, properties and irradiation experience are presented

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

The Feasibility of Administering a Practical Clinical Examination in Podiatry at a College of Podiatric Medicine: Results of a Field Trial Under Simulated Part III Test Conditions.

Science.gov (United States)

And Others; Valletta, Michael

1978-01-01

The results of a practical clinical examination in podiatric medicine administered to fourth-year students are presented. The examination could become the prototype of a Part III practical clinical examination under the auspices of the National Board of Podiatry Examiners. Its feasibility is established and problems and issues are discussed.…

The gamma spectrometry a powerful tool for irradiated fuel and fission products release studies

International Nuclear Information System (INIS)

Pontillon, Y.; Roure, C.; Lacroix, B.; Martella, T.; Ducros, G.; Ravel, S.; Gleizes, B.

2003-01-01

Over the last decades, due to the potentially severe consequences of a nuclear incident and/or accident for surrounding populations as well as the environment, international safety authorities launched R and D programs in support of general policy on exploitation of nuclear energy. This increasing interest enabled starting of many research programs in CEA and particularly in Nuclear Energy Directorate (DEN). Most of them are devoted to (i) the source term of fission products (including gas) and actinides released from PWR fuel samples in normal or accident conditions, (ii) burn-up determination, (iii) isotopic repartition... by quantitative gamma spectrometry. In this context, the Department of Fuel Studies (DEC), part of the DEN, has acquired considerable experience in this field of research. In order to attain the required capabilities, specific technical facilities set up in shielded hot cells at the CEA-Grenoble and CEA-Cadarache have been developed. In particular, the researchers of the Department have developed several gamma scanning benches and a set of two thermal treatment devices, including the so-called 'VERCORS facility'. These devices are associated to on line quantitative gamma spectrometry, in order to measure emitted gas and fission products (FPs). The greatest asset of such installations is to ensure a high analytical experiments rate, and as a consequence to make parametrical approach of planned studies easier. The first part of the present communication focuses, on the one hand, on the peculiar aspects of the gamma spectrometry applied on irradiated fuel, mad on the other hand, on the technical aspect of the different facilities (i.e. quantitative gamma spectrometry apparatus and corresponding 'home made' software). The last part is devoted to the results which can be obtained with such installation. In particular, it will be explained how experimental programs on FPs and gas release in normal and/or accidental conditions can be conducted

Alternative Fuels in Cement Production

DEFF Research Database (Denmark)

Larsen, Morten Boberg

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

Formation constants of Sm(III), Dy(III), Gd(III), Pr(III) and Nd(III) complexes of tridentate schiff base, 2-[(1H-benzimidazol-2-yl-methylene) amino] phenol

International Nuclear Information System (INIS)

Omprakash, K.L.; Chandra Pal, A.V.; Reddy, M.L.N.

1982-01-01

A new tridentate schiff base, 2- (1H-benzimidazol-2-yl-methylene)amino phenol derived from benzimididazole-2-carbo-xaldehyde and 2-aminophenol has been synthesised and characterised by spectral and analytical data. Proton-ligand formation constants of the schiff base and metal-ligand formation constants of its complexes with Sm(III), Dy(III), Gd(III), Nd(III) and Pr(III) have been determined potentiometrically in 50% (v/v) aqueous dioxane at an ionic strength of 0.1M (NaClO 4 ) and at 25deg C using the Irving-Rossotti titration technique. The order of stability constants (logβ 2 ) is found to be Sm(III)>Dy(III)>Gd(III)>Pr(III)>Nd(III). (author)

Formation constants of Sm(III), Dy(III), Gd(III), Pr(III) and Nd(III) complexes of tridentate schiff base, 2-((1H-benzimidazol-2-yl-methylene) amino) phenol

Energy Technology Data Exchange (ETDEWEB)

Omprakash, K L; Chandra Pal, A V; Reddy, M L.N. [Osmania Univ., Hyderabad (India). Dept. of Chemistry

1982-03-01

A new tridentate schiff base, 2- (1H-benzimidazol-2-yl-methylene)amino phenol derived from benzimididazole-2-carbo-xaldehyde and 2-aminophenol has been synthesised and characterised by spectral and analytical data. Proton-ligand formation constants of the schiff base and metal-ligand formation constants of its complexes with Sm(III), Dy(III), Gd(III), Nd(III) and Pr(III) have been determined potentiometrically in 50% (v/v) aqueous dioxane at an ionic strength of 0.1M (NaClO/sub 4/) and at 25deg C using the Irving-Rossotti titration technique. The order of stability constants (log..beta../sub 2/) is found to be Sm(III)>Dy(III)>Gd(III)>Pr(III)>Nd(III).

Drilling miniature holes, Part III

Energy Technology Data Exchange (ETDEWEB)

Gillespie, L.K.

1978-07-01

Miniature components for precision electromechanical mechanisms such as switches, timers, and actuators typically require a number of small holes. Because of the precision required, the workpiece materials, and the geometry of the parts, most of these holes must be produced by conventional drilling techniques. The use of such techniques is tedious and often requires considerable trial and error to prevent drill breakage, minimize hole mislocation and variations in hole diameter. This study of eight commercial drill designs revealed that printed circuit board drills produced better locational and size repeatability than did other drills when centerdrilling was not used. Boring holes 1 mm in dia, or less, as a general rule did not improve hole location in brass or stainless steel. Hole locations of patterns of 0.66-mm holes can be maintained within 25.4-..mu..m diametral positional tolerance if setup misalignments can be eliminated. Size tolerances of +- 3.8 ..mu..m can be maintained under some conditions when drilling flat plates. While these levels of precision are possible with existing off-the-shelf drills, they may not be practical in many cases.

Manufacturing and Construction of Spent Fuel Storage Rack for Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Sangjin; Jung, Kwangsub; Oh, Jinho; Lee, Jongmin [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The spent fuel storage rack consists of spent fuel storage racks and support frame. The spent fuel storage racks are installed in the support frame. A spent fuel storage rack consists of frame weldment and storage cell pipe assembly. Storage cell pipe assembly is mounted on the base plate of the frame weldment. The spent fuel storage rack is designed to withstand seismic load and other loads during earthquake. The structural integrity of the spent fuel storage rack is evaluated in accordance with ASME Section III, Subsection NF. Computer Code used for this analysis is ANSYS version 14.0.0. Dead load and seismic load is considered in load condition and hydrodynamic mass is included in the analysis. Design, manufacturing, and construction of the spent fuel storage rack are introduced. The spent fuel storage rack is for storage of spent fuel assemblies. The spent fuel storage rack should be designed, manufactured, and installed with consideration of predicted number of spent fuel assemblies, structural integrity, resistivity to corrosion and radiation, cleaning, and workability.

Reference Neutron Radiographs of Nuclear Reactor Fuel

DEFF Research Database (Denmark)

Domanus, Joseph Czeslaw

1986-01-01

Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

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

Science.gov (United States)

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

2014-03-01

Fuels of the furan family, i.e. furan itself, 2-methylfuran (MF), and 2,5-dimethylfuran (DMF) are being proposed as alternatives to hydrocarbon fuels and are potentially accessible from cellulosic biomass. While some experiments and modeling results are becoming available for each of these fuels, a comprehensive experimental and modeling analysis of the three fuels under the same conditions, simulated using the same chemical reaction model, has - to the best of our knowledge - not been attempted before. The present series of three papers, detailing the results obtained in flat flames for each of the three fuels separately, reports experimental data and explores their combustion chemistry using kinetic modeling. The first part of this series focuses on the chemistry of low-pressure furan flames. Two laminar premixed low-pressure (20 and 40 mbar) flat argon-diluted (50%) flames of furan were studied at two equivalence ratios (φ=1.0 and 1.7) using an analytical combination of high-resolution electron-ionization molecular-beam mass spectrometry (EI-MBMS) in Bielefeld and gas chromatography (GC) in Nancy. The time-of-flight MBMS with its high mass resolution enables the detection of both stable and reactive species, while the gas chromatograph permits the separation of isomers. Mole fractions of reactants, products, and stable and radical intermediates were measured as a function of the distance to the burner. A single kinetic model was used to predict the flame structure of the three fuels: furan (in this paper), 2-methylfuran (in Part II), and 2,5-dimethylfuran (in Part III). A refined sub-mechanism for furan combustion, based on the work of Tian et al. [Combustion and Flame 158 (2011) 756-773] was developed which was then compared to the present experimental results. Overall, the agreement is encouraging. The main reaction pathways involved in furan combustion were delineated computing the rates of formation and consumption of all species. It is seen that the

Technical Reports (Part I). End of Project Report, 1968-1971, Volume III.

Science.gov (United States)

Western Nevada Regional Education Center, Lovelock.

The pamphlets included in this volume are technical reports prepared as outgrowths of the Student Information Systems of the Western Nevada Regional Education Center (WN-REC) funded by a Title III (Elementary and Secondary Education Act) grant. These reports describe methods of interpreting the printouts from the Student Information System;…

Spectrophotometric and pH-Metric Studies of Ce(III, Dy(III, Gd(III,Yb(III and Pr(III Metal Complexes with Rifampicin

Directory of Open Access Journals (Sweden)

A. N. Sonar

2011-01-01

Full Text Available The metal-ligand and proton-ligand stability constant of Ce(III, Dy(III, Gd(III,Yb(III and Pr(III metals with substituted heterocyclic drug (Rifampicin were determined at various ionic strength by pH metric titration. NaClO4 was used to maintain ionic strength of solution. The results obtained were extrapolated to the zero ionic strength using an equation with one individual parameter. The thermodynamic stability constant of the complexes were also calculated. The formation of complexes has been studied by Job’s method. The results obtained were of stability constants by pH metric method is confirmed by Job’s method.

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

Complexes of lanthanum(III), cerium(III), samarium(III) and dysprosium(III) with substituted piperidines

Energy Technology Data Exchange (ETDEWEB)

Manhas, B S; Trikha, A K; Singh, H; Chander, M

1983-11-01

Complexes of the general formulae M/sub 2/Cl/sub 6/(L)/sub 3/.C/sub 2/H/sub 5/OH and M/sub 2/(NO/sub 3/)/sub 6/(L)/sub 2/.CH/sub 3/OH have been synthesised by the reactions of chlorides and nitrates of La(III), Ce(III), Sm(III) and Dy(III) with 2-methylpiperidine, 3-methylpiperidine and 4-methylpiperidine. These complexes have been characterised on the basis of their elemental analysis, and IR and electronic reflectance spectra. IR spectral data indicate the presence of coordinated ethanol and methanol molecules and bidentate nitrate groups. Coordination numbers of the metal ions vary from 5 to 8. 19 refs.

Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

International Nuclear Information System (INIS)

Wang Shisheng; Rineiski, Andrei; Maschek, Werner

2006-01-01

Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF 2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF 3 , etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as

Spectrophotometric determination of plutonium with chlorophosphonazo III in n-pentanol

International Nuclear Information System (INIS)

Saponara, N.M.; Marsh, S.F.

1982-03-01

Microgram amounts of plutonium are measured spectrophotometrically as the plutonium-chlorophosphonazo III complex after extraction into n-pentanol from 1.5 M HCl. The relative standard deviation is 1.5% for the range of 2.5 to 17.5 μg. The tolerance is excellent for many metals and nonmetals present in nuclear fuel-cycle materials. A preceding anion-exchange-column separation increases tolerance for certain metals and nonmetals

Direct methanol feed fuel cell and system

Science.gov (United States)

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

2009-01-01

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

A-Part Gel, an adhesion prophylaxis for abdominal surgery: a randomized controlled phase I-II safety study [NCT00646412].

Science.gov (United States)

Lang, Reinhold; Baumann, Petra; Schmoor, Claudia; Odermatt, Erich K; Wente, Moritz N; Jauch, Karl-Walter

2015-01-01

Intra-abdominal surgical intervention can cause the development of intra-peritoneal adhesions. To reduce this problem, different agents have been tested to minimize abdominal adhesions; however, the optimal adhesion prophylaxis has not been found so far. Therefore, the A-Part(®) Gel was developed as a barrier to diminish postsurgical adhesions; the aim of this randomized controlled study was a first evaluation of its safety and efficacy. In this prospective, controlled, randomized, patient-blinded, monocenter phase I-II study, 62 patients received either the hydrogel A-Part-Gel(®) as an anti-adhesive barrier or were untreated after primary elective median laparotomy. Primary endpoint was the occurrence of peritonitis and/or wound healing impairment 28 ± 10 days postoperatively. As secondary endpoints anastomotic leakage until 28 days after surgery, adverse events and adhesions were assessed until 3 months postoperatively. A lower rate of wound healing impairment and/or peritonitis was observed in the A-Part Gel(®) group compared to the control group: (6.5 vs. 13.8 %). The difference between the two groups was -7.3%, 90 % confidence interval [-20.1, 5.4 %]. Both treatment groups showed similar frequency of anastomotic leakage but incidence of adverse events and serious adverse events were slightly lower in the A-Part Gel(®) group compared to the control. Adhesion rates were comparable in both groups. A-Part Gel(®) is safe as an adhesion prophylaxis after abdominal wall surgery but no reduction of postoperative peritoneal adhesion could be found in comparison to the control group. This may at least in part be due to the small sample size as well as to the incomplete coverage of the incision due to the used application. NCT00646412.

WWER-440 fuel cycles possibilities using improved fuel assemblies design

International Nuclear Information System (INIS)

Mikolas, P.; Svarny, J.

2008-01-01

Practically five years cycle has been achieved in the last years at NPP Dukovany. There are two principal means how it could be achieved. First, it is necessary to use fuel assemblies with higher fuel enrichment and second, to use fuel loading with very low leakage. Both these conditions are fulfilled at NPP Dukovany at this time. It is known, that the fuel cycle economy can be improved by increasing the fuel residence time in the core up to six years. There are at least two ways how this goal could be achieved. The simplest way is to increase enrichment in fuel. There exists a limit, which is 5.0 w % of 235 U. Taking into account some uncertainty, the calculation maximum is 4.95 w % of 235 U. The second way is to change fuel assembly design. There are several possibilities, which seem to be suitable from the neutron - physical point of view. The first one is higher mass content of uranium in a fuel assembly. The next possibility is to enlarge pin pitch. The last possibility is to 'omit' FA shroud. This is practically unrealistic; anyway, some other structural parts must be introduced. The basic neutron physical characteristics of these cycles for up-rated power are presented showing that the possibilities of fuel assemblies with this improved design in enlargement of fuel cycles are very promising. In the end, on the basis of neutron physical characteristics and necessary economical input parameters, a preliminary evaluation of economic contribution of proposals of advanced fuel assemblies on fuel cycle economy is presented (Authors)

Fuels planning: science synthesis and integration; social issues fact sheet 13: Strategies for managing fuels and visual quality

Science.gov (United States)

Christine Esposito

2006-01-01

The public's acceptance of forest management practices, including fuels reduction, is heavily based on how forests look. Fuels managers can improve their chances of success by considering aesthetics when making management decisions. This fact sheet reviews a three-part general strategy for managing fuels and visual quality: planning, implementation, and monitoring...

HTPEM Fuel Cell Impedance

DEFF Research Database (Denmark)

Vang, Jakob Rabjerg

As part of the process to create a fossil free Denmark by 2050, there is a need for the development of new energy technologies with higher efficiencies than the current technologies. Fuel cells, that can generate electricity at higher efficiencies than conventional combustion engines, can...... potentially play an important role in the energy system of the future. One of the fuel cell technologies, that receives much attention from the Danish scientific community is high temperature proton exchange membrane (HTPEM) fuel cells based on polybenzimidazole (PBI) with phosphoric acid as proton conductor....... This type of fuel cell operates at higher temperature than comparable fuel cell types and they distinguish themselves by high CO tolerance. Platinum based catalysts have their efficiency reduced by CO and the effect is more pronounced at low temperature. This Ph.D. Thesis investigates this type of fuel...

Inner-sphere and outer-sphere complexes of yttrium(III), lanthanum (III), neodymium(III), terbium(III) and thulium(III) with halide ions in N,N-dimethylformamide

International Nuclear Information System (INIS)

Takahashi, Ryouta; Ishiguro, Shin-ichi

1991-01-01

The formation of chloro, bromo and iodo complexes of yttrium(III), and bromo and iodo complexes of lanthanum(III), neodymium(III), terbium(III) and thulium(III) has been studied by precise titration calorimetry in N,N-dimethylformamide (DMF) at 25 o C. The formation of [YCl] 2+ , [YCl 2 ] + , [YCl 3 ] and [YCl 4 ] - , and [MBr] 2+ and [MBr 2 ] + (M = Y, La, Nd, Tb, Tm) was revealed, and their formation constants, enthalpies and entropies were determined. It is found that the formation enthalpies change in the sequence ΔH o (Cl) > ΔH o (l), which is unusual for hard metal (III) ions. This implies that, unlike the chloride ion, the bromide ion forms outer-sphere complexes with the lanthanide(III) and yttrium(III) ions in DMF. Evidence for either an inner- or outer-sphere complex was obtained from 89 Y NMR spectra for Y(ClO 4 ) 3 , YCl 3 and YBr 3 DMF solutions at room temperature. (author)

LIFE Materials: Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward Volume 3

Energy Technology Data Exchange (ETDEWEB)

Turchi, P A; Kaufman, L; Fluss, M

2008-12-19

The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical, and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report (Volume 8 - Molten-salt Fuels) is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermo-chemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenges are not insurmountable, and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

Method for repairing failed fuel

International Nuclear Information System (INIS)

Shakudo, Taketomi.

1986-01-01

Purpose: To repair fuel elements that became failed during burnup in a reactor or during handling. Method: After the surface in the vicinity of a failed part of a fuel element is cleaned, a socket made of a shape-memory alloy having a ring form or a horseshoe form made by cutting a part of the ring form is inserted into the failed position according to the position of the failed fuel element. The shape memory alloy socket remembers a slightly larger inside diameter in its original phase (high-temperature side) than the outside diameter of the cladding tube and also a slightly larger inside diameter of the socket in the martensite phase (low-temperature side) than the outside diameter of the cladding tube, such that the socket can easily be inserted into the failed position. The socket, inserted into the failed part of the cladding tube, is heated by a heating jig. The socket recovers the original phase, and the shape also tends to recover a smaller diameter than the outside diameter of the cladding tube that has been remembered, and accordingly the failed part of the cladding tube is fastened with a great force and the failed part is fully closed with the socket, thus keeping radioactive materials from going out. (Horiuchi, T.)

Fire and blast safety manual for fuel element manufacture

International Nuclear Information System (INIS)

Ensinger, U.; Koehler, B.; Mester, W.; Riotte, H.G.; Sehrbrock, H.W.

1988-01-01

The manual aims to enable people involved in the planning, operation, supervision, licensing or appraisal of fuel element factories to make a quick and accurate assessment of blast safety. In Part A, technical plant principles are shown, and a summary lists the flammable materials and ignition sources to be found in fuel element factories, together with theoretical details of what happens during a fire or a blast. Part B comprises a list of possible fires and explosions in fuel element factories and ways of preventing them. Typical fire and explosion scenarios are analysed more closely on the basis of experiments. Part B also contains a list and an assessment of actual fires and explosions which have occurred in fuel element factories. Part C contains safety measures to protect against fire and explosion, in-built fire safety, fire safety in plant design, explosion protection and measures to protect people from radiation and other hazards when fighting fires. A distinction is drawn between UO 2 , MOX and HTR fuel elements. (orig./DG) [de

Licensed fuel facility status report: Inventory difference data, January 1988--June 1988

International Nuclear Information System (INIS)

1989-03-01

NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, after Agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than (i) one effective kilogram of special nuclear material of low strategic significance, (ii) one effective kilogram of special nuclear material of moderate strategic significance, (iii) one effective kilogram of strategic special nuclear material contained in irradiated fuel reprocessing operations, or (iv) five formula kilograms of strategic special nuclear material

Accuracy of fuel motion measurements using in-core detectors

International Nuclear Information System (INIS)

Dupree, S.A.

1975-01-01

An initial assessment has been made as to how accurately fuel motion can be measured with in-core detectors. A portion of this assessment has involved the calculation of the response of various detectors to fuel motion and the development of a formalism for correlating uncertainties in a neutron flux measurement to uncertainties in the fuel motion. Initially, four idealized configurations were studied in one dimension. These configurations consisted of (1) a single fuel-pin test using ACPR, (2) a seven fuel-pin test using ACPR, (3) a full subassembly (271 pin) test using a Class I ANL-type SAREF, and (4) a full subassembly plus six partial subassemblies (approximately 1000 pin) test using a Class III GE-type SAREF. It was assumed that melt would occur symmetrically at the center of the test fuel and that fuel would therefore disappear from the center of the geometry. For each case of series of calculations was performed in which detector responses were determined at several radial locations for the unperturbed core and for the core with various fractions of the fuel replaced with Na. This fuel loss was assumed to occur essentially instantaneously such that the power level in the remaining portion of the test fuel remained unchanged from that of the initial unperturbed condition

Alcohol fuels for developing countries

International Nuclear Information System (INIS)

Bhattacharya, Partha

1993-01-01

The importance of alcohol as an alternative fuel has been slowly established. In countries such as Brazil, they are already used in transport and other sectors of economy. Other developing countries are also trying out experiments with alcohol fuels. Chances of improving the economy of many developing nations depends to a large extent on the application of this fuel. The potential for alcohol fuels in developing countries should be considered as part of a general biomass-use strategy. The final strategies for the development of alcohol fuel will necessarily reflect the needs, values, and conditions of the individual nations, regions, and societies that develop them. (author). 5 refs

The European carbon balance. Part 1: fossil fuel emissions

NARCIS (Netherlands)

Ciais, P.; Paris, J.D.; Marland, G.; Peylin, P.; Piao, S.L.; levin, I.; Pregger, T.; Scholz, Y.; Friedrich, R.; Rivier, L.; Houweling, S.; Schulze, E.D.

2010-01-01

We analyzed the magnitude, the trends and the uncertainties of fossil-fuel CO2 emissions in the European Union 25 member states (hereafter EU-25), based on emission inventories from energy-use statistics. The stability of emissions during the past decade at EU-25 scale masks decreasing trends in

Utility residual fuel oil market conditions: An update

International Nuclear Information System (INIS)

Mueller, H.A. Jr.

1992-01-01

Planning for residual fuel oil usage and management remains an important part of the generation fuel planning and management function for many utilities. EPRI's Utility Planning Methods Center has maintained its analytical overview of the fuel oil markets as part of its overall fuel planning and management research program. This overview provides an update of recent fuel oil market directions. Several key events of the past year have had important implications for residual fuel oil markets. The key events have been the changes brought about by the Persian Gulf War and its aftermath, as well as continuing environmental policy developments. The Persian Gulf conflict has created renewed interest in reducing fuel oil use by utilities as part of an overall reduction in oil imports. The policy analysis performed to date has generally failed to properly evaluate utility industry capability. The Persian Gulf conflict has also resulted in an important change in the structure of international oil markets. The result of this policy-based change is likely to be a shift in oil pricing strategy. Finally, continued change in environmental requirements is continuing to shift utility residual oil requirements, but is also changing the nature of the US resid market itself

Sulphur release from alternative fuel firing

DEFF Research Database (Denmark)

Cortada Mut, Maria del Mar; Nørskov, Linda Kaare; Glarborg, Peter

2014-01-01

The cement industry has long been dependent on the use of fossil fuels, although a recent trend in replacing fossil fuels with alternative fuels has arisen. 1, 2 However, when unconverted or partly converted alternative fuels are admitted directly in the rotary kiln inlet, the volatiles released...... from the fuels may react with sulphates present in the hot meal to form SO 2 . Here Maria del Mar Cortada Mut and associates describe pilot and industrial scale experiments focusing on the factors that affect SO 2 release in the cement kiln inlet....

Overview of chemical characterization of FBTR fuel

International Nuclear Information System (INIS)

Venkatesan, V.; Nandi, C.; Patil, A.B.; Prakash, Amrit; Khan, K.B.; Arun Kumar

2015-01-01

Uranium Plutonium mixed carbide fuel is the driver fuel for Fast Breeder Test Reactor (FBTR) at IGCAR. The fuel is being fabricated at Radiometallurgy Division, BARC by conventional powder metallurgy route. During the fabrication of fuel, chemical quality control of process intermediates is very important to reach stringent specification of the final fuel product. Different steps are involved in the fabrication of uranium-plutonium carbide (MC) for FBTR. The main steps in the fabrication of MC fuel pellets are carbothermic reduction (CR) of mixture of uranium oxide, plutonium oxide and graphite powder to prepare MC clinkers, crushing and milling of MC clinkers and consolidation of MC powders into fuel pellets and sintering. As a part of process control, analysis of uranium (U), plutonium (Pu), carbon in oxide graphite mixture and U, Pu, carbon, oxygen, nitrogen, MC, M 2 C 3 contents in mixed carbide powder (MC clinkers) are carried out at our laboratory. Analysis of U, Pu, carbon, oxygen, nitrogen, MC and M 2 C 3 contents in mixed carbide sintered pellets are carried out as a part of quality control. This paper describes an overview of analytical instruments used during chemical quality control of mixed carbide fuel

The Bacterial Flagellar Type III Export Gate Complex Is a Dual Fuel Engine That Can Use Both H+ and Na+ for Flagellar Protein Export.

Directory of Open Access Journals (Sweden)

Tohru Minamino

2016-03-01

Full Text Available The bacterial flagellar type III export apparatus utilizes ATP and proton motive force (PMF to transport flagellar proteins to the distal end of the growing flagellar structure for self-assembly. The transmembrane export gate complex is a H+-protein antiporter, of which activity is greatly augmented by an associated cytoplasmic ATPase complex. Here, we report that the export gate complex can use sodium motive force (SMF in addition to PMF across the cytoplasmic membrane to drive protein export. Protein export was considerably reduced in the absence of the ATPase complex and a pH gradient across the membrane, but Na+ increased it dramatically. Phenamil, a blocker of Na+ translocation, inhibited protein export. Overexpression of FlhA increased the intracellular Na+ concentration in the presence of 100 mM NaCl but not in its absence, suggesting that FlhA acts as a Na+ channel. In wild-type cells, however, neither Na+ nor phenamil affected protein export, indicating that the Na+ channel activity of FlhA is suppressed by the ATPase complex. We propose that the export gate by itself is a dual fuel engine that uses both PMF and SMF for protein export and that the ATPase complex switches this dual fuel engine into a PMF-driven export machinery to become much more robust against environmental changes in external pH and Na+ concentration.

The Bacterial Flagellar Type III Export Gate Complex Is a Dual Fuel Engine That Can Use Both H+ and Na+ for Flagellar Protein Export.

Science.gov (United States)

Minamino, Tohru; Morimoto, Yusuke V; Hara, Noritaka; Aldridge, Phillip D; Namba, Keiichi

2016-03-01

The bacterial flagellar type III export apparatus utilizes ATP and proton motive force (PMF) to transport flagellar proteins to the distal end of the growing flagellar structure for self-assembly. The transmembrane export gate complex is a H+-protein antiporter, of which activity is greatly augmented by an associated cytoplasmic ATPase complex. Here, we report that the export gate complex can use sodium motive force (SMF) in addition to PMF across the cytoplasmic membrane to drive protein export. Protein export was considerably reduced in the absence of the ATPase complex and a pH gradient across the membrane, but Na+ increased it dramatically. Phenamil, a blocker of Na+ translocation, inhibited protein export. Overexpression of FlhA increased the intracellular Na+ concentration in the presence of 100 mM NaCl but not in its absence, suggesting that FlhA acts as a Na+ channel. In wild-type cells, however, neither Na+ nor phenamil affected protein export, indicating that the Na+ channel activity of FlhA is suppressed by the ATPase complex. We propose that the export gate by itself is a dual fuel engine that uses both PMF and SMF for protein export and that the ATPase complex switches this dual fuel engine into a PMF-driven export machinery to become much more robust against environmental changes in external pH and Na+ concentration.

Winter fuels report

Energy Technology Data Exchange (ETDEWEB)

1995-01-13

The Winter Fuels Report is intended to provide concise, timely information to the industry, the press, policymakers, consumers, analysts, and State and local governments on the following topics: distillate fuel oil net production, imports and stocks on a US level and for all Petroleum Administration for Defense Districts (PADD) and product supplied on a US level; propane net production, imports and stocks on a US level and for PADD`s I, II, and III; natural gas supply and disposition and underground storage for the US and consumption for all PADD`s, as well as selected National average prices; residential and wholesale pricing data for heating oil and propane for those States participating in the joint Energy Information Administration (EIA)/State Heating Oil and Propane Program; crude oil and petroleum price comparisons for the US and selected cities; and a 6-10 day, 30-Day, and 90-Day outlook for temperature and precipitation and US total heating degree-days by city.

Winter fuels report

International Nuclear Information System (INIS)

1995-01-01

The Winter Fuels Report is intended to provide concise, timely information to the industry, the press, policymakers, consumers, analysts, and State and local governments on the following topics: distillate fuel oil net production, imports and stocks on a US level and for all Petroleum Administration for Defense Districts (PADD) and product supplied on a US level; propane net production, imports and stocks on a US level and for PADD's I, II, and III; natural gas supply and disposition and underground storage for the US and consumption for all PADD's, as well as selected National average prices; residential and wholesale pricing data for heating oil and propane for those States participating in the joint Energy Information Administration (EIA)/State Heating Oil and Propane Program; crude oil and petroleum price comparisons for the US and selected cities; and a 6-10 day, 30-Day, and 90-Day outlook for temperature and precipitation and US total heating degree-days by city

LPG diesel dual fuel engine – A critical review

Directory of Open Access Journals (Sweden)

B. Ashok

2015-06-01

Full Text Available The engine, which uses both conventional diesel fuel and LPG fuel, is referred to as ‘LPG–diesel dual fuel engines’. LPG dual fuel engines are modified diesel engines which use primary fuel as LPG and secondary fuel as diesel. LPG dual fuel engines have a good thermal efficiency at high output but the performance is less during part load conditions due to the poor utilization of charges. This problem can be overcome by varying factors such as pilot fuel quantity, injection timing, composition of the gaseous fuel and intake charge conditions, for improving the performance, combustion and emissions of dual fuel engines. This article reviews about the research work done by the researchers in order to improve the performance, combustion and emission parameters of a LPG–diesel dual fuel engines. From the studies it is shown that the use of LPG in diesel engine is one of the capable methods to reduce the PM and NOx emissions but at same time at part load condition there is a drop in efficiency and power output with respect to diesel operation.

Fuel rod fixing system

International Nuclear Information System (INIS)

Christiansen, D.W.

1982-01-01

This is a reusable system for fixing a nuclear reactor fuel rod to a support. An interlock cap is fixed to the fuel rod and an interlock strip is fixed to the support. The interlock cap has two opposed fingers, which are shaped so that a base is formed with a body part. The interlock strip has an extension, which is shaped so that this is rigidly fixed to the body part of the base. The fingers of the interlock cap are elastic in bending. To fix it, the interlock cap is pushed longitudinally on to the interlock strip, which causes the extension to bend the fingers open in order to engage with the body part of the base. To remove it, the procedure is reversed. (orig.) [de

Liquid fuel concept benefits

International Nuclear Information System (INIS)

Hron, M.

1996-01-01

There are principle drawbacks of any kind of solid nuclear fuel listed and analyzed in the first part of the paper. One of the primary results of the analyses performed shows that the solid fuel concept, which was to certain degree advantageous in the first periods of a nuclear reactor development and operation, has guided this branch of a utilization of atomic nucleus energy to a death end. On the background of this, the liquid fuel concept and its benefits are introduced and briefly described in the first part of the paper, too. As one of the first realistic attempts to utilize the advantages of liquid fuels, the reactor/blanket system with molten fluoride salts in the role of fuel and coolant simultaneously, as incorporated in the accelerator-driven transmutation technology (ADTT) being proposed and currently having been under development in the Los Alamos National Laboratory, will be studied both theoretically and experimentally. There is a preliminary design concept of an experimental assembly LA-O briefly introduced in the paper which is under preparation in the Czech Republic for such a project. Finally, there will be another very promising concept of a small low power ADTT system introduced which is characterized by a high level of safety and economical efficiency. In the conclusion, the overall survey of principal benefits which may be expected by introducing liquid nuclear fuel in nuclear power and research reactor systems is given and critically analyzed. 7 refs, 4 figs

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.

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part I

International Nuclear Information System (INIS)

2011-01-01

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

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.

Physiotherapy and low back pain - part iii: outcomes research utilising the biosychosocial model: psychosocial outcomes

Directory of Open Access Journals (Sweden)

L. D. Bardin

2003-02-01

has evolved that necessitates the use of a biopsychosocial model, focusing on illness rather than disease and incorporating the biological, psychological and social aspects that are important to understand and to study LBP in its chronic form. Traditional outcome measures that measure elements within the biological component are limited to assess the spectrum of impacts caused by chronic low back pain (CLBP and the validity, reliability and sensitivity of some of these measures has been questioned.Few physiologic tests of spine function are clinically meaningful to patients, objective physical findings can be absent, and in CLBP disability and activity intolerance are often disproportional to the original injury. Biological outcomes should be complemented by outcomes of the psychosocial aspects of back pain that measure the considerable functional and emotional impact on the quality of life of patients experiencing low back dysfunction. Outcomes research is an analysis of clinical practice as it actually occurs and can  make a valuable contribution to understanding the multidimensional impact of LBP. Psychosocial aspects of the biopsychosocial model for outcomes research are discussed in part III: functional status/disability, psychological impairment, patient satisfaction, health related quality of life

Toxic and hazardous chemicals, Title III and communities: An outreach manual for community groups