WorldWideScience

Sample records for distillate test fuels

  1. Harmonisation of microbial sampling and testing methods for distillate fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hill, G.C.; Hill, E.C. [ECHA Microbiology Ltd., Cardiff (United Kingdom)

    1995-05-01

    Increased incidence of microbial infection in distillate fuels has led to a demand for organisations such as the Institute of Petroleum to propose standards for microbiological quality, based on numbers of viable microbial colony forming units. Variations in quality requirements, and in the spoilage significance of contaminating microbes plus a tendency for temporal and spatial changes in the distribution of microbes, makes such standards difficult to implement. The problem is compounded by a diversity in the procedures employed for sampling and testing for microbial contamination and in the interpretation of the data obtained. The following paper reviews these problems and describes the efforts of The Institute of Petroleum Microbiology Fuels Group to address these issues and in particular to bring about harmonisation of sampling and testing methods. The benefits and drawbacks of available test methods, both laboratory based and on-site, are discussed.

  2. One-day stability test for distillate fuel oils

    Energy Technology Data Exchange (ETDEWEB)

    Gyrath, F W; Dunn, Jr, F R; Smith, Jr, A C

    1958-08-01

    A one-day stability test is described. One liter of oil was placed in a glass bottle along with several steel strips, the bottle was purged with oxygen, sealed, and placed in an oven at 100/sup 0/C (212/sup 0/F) for 24 hours. At the end of the test, the oil was cooled and filtered, and the sediment on the filter was dried and weighed. The results of the one-day test were correlated with the results of storage tests in which oil samples were stored for six months at 29.4/sup 0/C (85/sup 0/F) in 30-gallon steel drums or in glass bottles containing steel strips. At the end of storage, sediment was measured by filtering and weighing. The one-day stability test was also compared with storage in 100-barrel tanks at ambient temperature. The one-day stability test was in good agreement with the storage tests, and could be used to predict the amount of sediment that would be formed in storage.

  3. 40 CFR 1065.703 - Distillate diesel fuel.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Distillate diesel fuel. 1065.703... Standards § 1065.703 Distillate diesel fuel. (a) Distillate diesel fuels for testing must be clean and... distillate diesel fuels: (1) Cetane improver. (2) Metal deactivator. (3) Antioxidant, dehazer. (4) Rust...

  4. Lubricity characteristics of marine distillate fuels

    Energy Technology Data Exchange (ETDEWEB)

    Crutchley, Ian [Innospec Fuel Specialties, Ellesmere Port (United Kingdom); Green, Michael [Intertek Lintec ShipCare Services, Darlington (United Kingdom)

    2012-08-15

    This article from Innospec Fuel Specialties, Ellesmere Port, UK, and Intertek Lintec ShipCare Services, Darlington, UK, examines the lubricity characteristics of marine distillate fuels available today in relation to the requirements and limits imposed in ISO8217:2010. It will estimate expected failure rates and also asses the perceived relationship between lubricity, sulphur content and viscosity. (orig.)

  5. Cascade Distillation Subsystem Development: Progress Toward a Distillation Comparison Test

    Science.gov (United States)

    Callahan, M. R.; Lubman, A.; Pickering, Karen D.

    2009-01-01

    Recovery of potable water from wastewater is essential for the success of long-duration manned missions to the Moon and Mars. Honeywell International and a team from NASA Johnson Space Center (JSC) are developing a wastewater processing subsystem that is based on centrifugal vacuum distillation. The wastewater processor, referred to as the Cascade Distillation Subsystem (CDS), utilizes an innovative and efficient multistage thermodynamic process to produce purified water. The rotary centrifugal design of the system also provides gas/liquid phase separation and liquid transport under microgravity conditions. A five-stage subsystem unit has been designed, built, delivered and integrated into the NASA JSC Advanced Water Recovery Systems Development Facility for performance testing. A major test objective of the project is to demonstrate the advancement of the CDS technology from the breadboard level to a subsystem level unit. An initial round of CDS performance testing was completed in fiscal year (FY) 2008. Based on FY08 testing, the system is now in development to support an Exploration Life Support (ELS) Project distillation comparison test expected to begin in early 2009. As part of the project objectives planned for FY09, the system will be reconfigured to support the ELS comparison test. The CDS will then be challenged with a series of human-gene-rated waste streams representative of those anticipated for a lunar outpost. This paper provides a description of the CDS technology, a status of the current project activities, and data on the system s performance to date.

  6. Cascade Distiller System Performance Testing Interim Results

    Science.gov (United States)

    Callahan, Michael R.; Pensinger, Stuart; Sargusingh, Miriam J.

    2014-01-01

    The Cascade Distillation System (CDS) is a rotary distillation system with potential for greater reliability and lower energy costs than existing distillation systems. Based upon the results of the 2009 distillation comparison test (DCT) and recommendations of the expert panel, the Advanced Exploration Systems (AES) Water Recovery Project (WRP) project advanced the technology by increasing reliability of the system through redesign of bearing assemblies and improved rotor dynamics. In addition, the project improved the CDS power efficiency by optimizing the thermoelectric heat pump (TeHP) and heat exchanger design. Testing at the NASA-JSC Advanced Exploration System Water Laboratory (AES Water Lab) using a prototype Cascade Distillation Subsystem (CDS) wastewater processor (Honeywell d International, Torrance, Calif.) with test support equipment and control system developed by Johnson Space Center was performed to evaluate performance of the system with the upgrades as compared to previous system performance. The system was challenged with Solution 1 from the NASA Exploration Life Support (ELS) distillation comparison testing performed in 2009. Solution 1 consisted of a mixed stream containing human-generated urine and humidity condensate. A secondary objective of this testing is to evaluate the performance of the CDS as compared to the state of the art Distillation Assembly (DA) used in the ISS Urine Processor Assembly (UPA). This was done by challenging the system with ISS analog waste streams. This paper details the results of the AES WRP CDS performance testing.

  7. Distillate Fuel Trends: International Supply Variations and Alternate Fuel Properties

    Science.gov (United States)

    2013-01-31

    fuel in NATO countries will have some amount of FAME present. There is some work being done on hydrocarbon alternatives but the regulatory structure ... synthesis or hydrotreatment – Requirements and test methods.” According to the specification, paraffinic diesel fuel does not meet the current requirements...or international specification for triglyceride based fuel oils (straight vegetable oil / raw vegetable oil). The same holds true for alcohol-based

  8. Low grade bioethanol for fuel mixing on gasoline engine using distillation process

    Science.gov (United States)

    Abikusna, Setia; Sugiarto, Bambang; Suntoro, Dedi; Azami

    2017-03-01

    Utilization of renewable energy in Indonesia is still low, compared to 34% oil, 20% coal and 20% gas, utilization of energy sources for water 3%, geothermal 1%, 2% biofuels, and biomass 20%. Whereas renewable energy sources dwindling due to the increasing consumption of gasoline as a fuel. It makes us have to look for alternative renewable energy, one of which is bio ethanol. Several studies on the use of ethanol was done to the researchers. Our studies using low grade bio ethanol which begins with the disitillation independently utilize flue gas heat at compact distillator, produces high grade bio ethanol and ready to be mixed with gasoline. Stages of our study is the compact distillator design of the motor dynamic continued with good performance and emission testing and ethanol distilled. Some improvement is made is through the flue gas heat control mechanism in compact distillator using gate valve, at low, medium, and high speed engine. Compact distillator used is kind of a batch distillation column. Column design process using the shortcut method, then carried the tray design to determine the overall geometry. The distillation is done by comparing the separator with a tray of different distances. As well as by varying the volume of the feed and ethanol levels that will feed distilled. In this study, we analyzed the mixing of ethanol through variation between main jet and pilot jet in the carburetor separately interchangeably with gasoline. And finally mixing mechanism bio ethanol with gasoline improved with fuel mixer for performance.

  9. Retort for distilling oil from shales and other fuels

    Energy Technology Data Exchange (ETDEWEB)

    Wood, E C

    1931-04-17

    A retort is constructed in the form of a flat shallow covered conduit through which the fuel in a pulverized form is continuously conveyed on a train of pans so loaded as to completely fill the conduit, heat being applied to an adequate central portion of the conduit to effect the distillation of the fuel. The roof or cover is sufficiently higher for the extent of such heated portion to afford space for and thereby facilitate the offtake of the distillation products while the ends of the conduit remain sealed against the escape of such products.

  10. 40 CFR 94.108 - Test fuels.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Test fuels. 94.108 Section 94.108... EMISSIONS FROM MARINE COMPRESSION-IGNITION ENGINES Test Procedures § 94.108 Test fuels. (a) Distillate diesel test fuel. (1) The diesel fuels for testing Category 1 and Category 2 marine engines designed to...

  11. Method of dry distillation of low-grade fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hellsing, G H; Wengstrom, R O.A.

    1920-05-20

    A method of dry distillation of low-grade fuels is characterized by having the process take place in a furnace that is charged alternately by partly cooled, red-hot, and fresh raw materials. The patent has one more claim.

  12. Sodium removal of fuel elements by vacuum distillation

    International Nuclear Information System (INIS)

    Buescher, E.; Haubold, W.; Jansing, W.; Kirchner, G.

    1978-01-01

    Cleaning of sodium-wetted core components can be performed by using either lead, moist nitrogen, or alcohol. The advantages of these methods for cleaning fuel elements without causing damage are well known. The disadvantage is that large amounts of radioactive liquids are formed during handling in the latter two cases. In this paper a new method to clean components is described. The main idea is to remove all liquid metal from the core components within a comparatively short period of time. Fuel elements removed from the reactor must be cooled because of high decay heat release. To date, vacuum distillation of fuel elements has not yet been applied

  13. Biodiesel of distilled hydrogenated fat and biodiesel of distilled residual oil: fuel consumption in agricultural tractor

    Energy Technology Data Exchange (ETDEWEB)

    Camara, Felipe Thomaz da; Lopes, Afonso; Silva, Rouverson Pereira da; Oliveira, Melina Cais Jejcic; Furlani, Carlos Eduardo Angeli [Universidade Estadual Paulista (UNESP), Jaboticabal, SP (Brazil); Dabdoub, Miguel Joaquim [Universidade de Sao Paulo (USP), Ribeirao Preto (Brazil)

    2008-07-01

    Great part of the world-wide oil production is used in fry process; however, after using, such product becomes an undesirable residue, and the usual methods of discarding of these residues, generally contaminate the environment, mainly the rivers. In function of this, using oil and residual fat for manufacturing biodiesel, besides preventing ambient contamination, turning up an undesirable residue in to fuel. The present work had as objective to evaluate the fuel consumption of a Valtra BM100 4x2 TDA tractor functioning with methylic biodiesel from distilled hydrogenated fat and methylic biodiesel from distilled residual oil, in seven blends into diesel. The work was conducted at the Department of Agricultural Engineering, at UNESP - Jaboticabal, in an entirely randomized block statistical design, factorial array of 2 x 7, with three repetitions. The factors combinations were two types of methylic distilled biodiesel (residual oil and hydrogenated fat) and seven blends (B{sub 0}, B{sub 5}, B{sub 1}5, B{sub 2}5, B{sub 5}0, B{sub 7}5 and B{sub 1}00). The results had evidenced that additioning 15% of biodiesel into diesel, the specific consumption was similar, and biodiesel of residual oil provided less consumption than biodiesel from hydrogenated fat. (author)

  14. Application of the Advanced Distillation Curve Method to Fuels for Advanced Combustion Engine Gasolines

    KAUST Repository

    Burger, Jessica L.

    2015-07-16

    © This article not subject to U.S. Copyright. Published 2015 by the American Chemical Society. Incremental but fundamental changes are currently being made to fuel composition and combustion strategies to diversify energy feedstocks, decrease pollution, and increase engine efficiency. The increase in parameter space (by having many variables in play simultaneously) makes it difficult at best to propose strategic changes to engine and fuel design by use of conventional build-and-test methodology. To make changes in the most time- and cost-effective manner, it is imperative that new computational tools and surrogate fuels are developed. Currently, sets of fuels are being characterized by industry groups, such as the Coordinating Research Council (CRC) and other entities, so that researchers in different laboratories have access to fuels with consistent properties. In this work, six gasolines (FACE A, C, F, G, I, and J) are characterized by the advanced distillation curve (ADC) method to determine the composition and enthalpy of combustion in various distillate volume fractions. Tracking the composition and enthalpy of distillate fractions provides valuable information for determining structure property relationships, and moreover, it provides the basis for the development of equations of state that can describe the thermodynamic properties of these complex mixtures and lead to development of surrogate fuels composed of major hydrocarbon classes found in target fuels.

  15. Cascade Distillation Subsystem Development: Early Results from the Exploration Life Support Distillation Technology Comparison Test

    Science.gov (United States)

    Callahan, Michael R.; Patel, Vipul; Pickering, Karen D.

    2010-01-01

    In 2009, the Cascade Distillation Subsystem (CDS) wastewater processor (Honeywell International, Torrance, California) was assessed in the National Aeronautics and Space Administration (NASA) Exploration Life Support (ELS) distillation comparison test. The purpose of the test was to collect data to support down-selection and development of a primary distillation technology for application in a lunar outpost water recovery system. The CDS portion of the comparison test was conducted between May 6 and August 19, 2009. The system was challenged with two pretreated test solutions, each intended to represent a feasible wastewater generated in a surface habitat. The 30-day equivalent wastewater loading volume for a crew of four was intended to be processed for each wastewater solution. Test Solution 1 consisted of a mixed stream containing human-generated urine and humidity condensate. Test Solution 2 contained the addition of human-generated hygiene wastewater to the solution 1 waste stream components. Approximately 1500 kg of total wastewater was processed through the CDS during testing. Respective recoveries per solution were 93.4 +/- 0.7 and 90.3 +/- 0.5 percent. The average specific energy of the system during testing was calculated to be less than 120 W-hr/kg. The following paper provides detailed information and data on the performance of the CDS as challenged per the ELS distillation comparison test.

  16. Honeywell Cascade Distiller System Performance Testing Interim Results

    Science.gov (United States)

    Callahan, Michael R.; Sargusingh, Miriam

    2014-01-01

    The ability to recover and purify water through physiochemical processes is crucial for realizing long-term human space missions, including both planetary habitation and space travel. Because of their robust nature, distillation systems have been actively pursued as one of the technologies for water recovery. The Cascade Distillation System (CDS) is a vacuum rotary distillation system with potential for greater reliability and lower energy costs than existing distillation systems. The CDS was previously under development through Honeywell and NASA. In 2009, an assessment was performed to collect data to support down-selection and development of a primary distillation technology for application in a lunar outpost water recovery system. Based on the results of this testing, an expert panel concluded that the CDS showed adequate development maturity, TRL-4, together with the best product water quality and competitive weight and power estimates to warrant further development. The Advanced Exploration Systems (AES) Water Recovery Project (WRP) worked to address weaknesses identified by The Panel; namely bearing design and heat pump power efficiency. Testing at the NASA-JSC Advanced Exploration System Water Laboratory (AES Water Lab) using a prototype Cascade Distillation Subsystem (CDS) wastewater processor (Honeywell International, Torrance, Calif.) with test support equipment and control system developed by Johnson Space Center was performed to evaluate performance of the system with the upgrades. The CDS will also have been challenged with ISS analog waste streams and a subset of those being considered for Exploration architectures. This paper details interim results of the AES WRP CDS performance testing.

  17. Importance of low-temperature distillation of coal for German fuel economics

    Energy Technology Data Exchange (ETDEWEB)

    Rosendahl, F

    1942-01-01

    Improved processes are available to give low-temperature distillation products economic importance. Low-temperature distillation is limited to the utilization of high-volatile nut coals and briquets. The coke formed can be used as a smokeless fuel, and the tar directly as a fuel oil. Phenols can be extracted, in order to work up the residue into fuel oil and motor fuel. Large deposits of coal in Upper Silesia and in the Saar District are suitable for low-temperature distillation.

  18. A rapid colorimetric method for predicting the storage stability of middle distillate fuels

    Energy Technology Data Exchange (ETDEWEB)

    Marshman, S.J. [Defense Research Agency, Surrey (United Kingdom)

    1995-05-01

    Present methods used to predict the storage stability of distillate fuels such as ASTM D2274, ASTM D4625, DEF STAN 05-50 Method 40 and in-house methods are very time consuming, taking a minimum of 16 hours. In addition, some of these methods under- or over-predict the storage stability of the test fuel. A rapid colorimetric test for identifying cracked, straight run or hydrofined fuels was reported at the previous Conference. Further work has shown that while a visual appraisal is acceptable for refinery-fresh fuels, colour development may be masked by other coloured compounds in older fuels. Use of a spectrometric finish to the method has extended the scope of the method to include older fuels. The test can be correlated with total sediment from ASTM D4625 (13 weeks at 43{degrees}C) over a sediment range of 0-60mg/L. A correlation of 0.94 was obtained for 40 fuels.

  19. Process of producing fuels from slates or bituminous shales. [distillation at incandescent heat

    Energy Technology Data Exchange (ETDEWEB)

    Huppenbauer, M

    1902-07-31

    A process of producing a fuel from slates or bituminous shales by saturating or impregnating them after preliminary distillation with the vapors of tars, resins, oils, etc., is given. The process is characterized by the bituminous shale being submitted in the form of fragments to distillation at incandescent heat to make the shale porous and able to absorb the vapors of the substances already mentioned.

  20. Method of dry distillation of solid and liquid fuels. [below 500/sup 0/C

    Energy Technology Data Exchange (ETDEWEB)

    Michot-Dupont, G F

    1932-07-29

    A method of manufacture is given of liquid fuels with a low boiling temperature by means of dry distillation of coals, lignites, peat, and similar materials at temperatures below 500/sup 0/C in the presence of fatty acids (such as acetates). It is characterized by carrying out the distillation after adding at least one more basic reacting salt in such quantity that the reaction takes place under highly basic conditions.

  1. Effect of hydroprocessing severity on characteristics of jet fuel from OSCO 2 and Paraho distillates

    Science.gov (United States)

    Prok, G. M.; Flores, F. J.; Seng, G. T.

    1981-01-01

    Jet A boiling range fuels and broad-property research fuels were produced by hydroprocessing shale oil distillates, and their properties were measured to characterize the fuels. The distillates were the fraction of whole shale oil boiling below 343 C from TOSCO 2 and Paraho syncrudes. The TOSCO 2 was hydroprocessed at medium severity, and the Paraho was hydroprocessed at high, medium, and low severities. Fuels meeting Jet A requirements except for the freezing point were produced from the medium severity TOSCO 2 and the high severity Paraho. Target properties of a broad property research fuel were met by the medium severity TOSCO 2 and the high severity Paraho except for the freezing point and a high hydrogen content. Medium and low severity Paraho jet fuels did not meet thermal stability and freezing point requirements.

  2. Emulsification of waste cooking oils and fatty acid distillates as diesel engine fuels: An attractive alternative

    Directory of Open Access Journals (Sweden)

    Eliezer Ahmed Melo Espinosa

    2016-06-01

    Full Text Available The scope of this paper is to analyze the possibility and feasibility of the use of emulsification method applied to waste cooking oils and fatty acid distillates as diesel engine fuels, compared with other commonly used methods. These waste products are obtained from the refining oil industry, food industry and service sector, mainly. They are rarely used as feedstock to produce biofuels and other things, in spite of constitute a potential source of environmental contamination. From the review of the state of arts, significant decreases in exhaust emissions of nitrogen oxides, cylinder pressure as well as increases of the ignition delay, brake specific fuel consumption, hydrocarbon, smoke opacity, carbon monoxide, particulate matters to emulsified waste cooking oils and fatty acid distillates compared with diesel fuel are reported. In some experiments the emulsified waste cooking oils achieved better performance than neat fatty acid distillates, neat waste cooking oils and their derivatives methyl esters.

  3. Ethanol dehydration via azeotropic distillation with gasoline fractions as entrainers: A pilot-scale study of the manufacture of an ethanol–hydrocarbon fuel blend

    OpenAIRE

    Gomis Yagües, Vicente; Pedraza Berenguer, Ricardo; Saquete Ferrándiz, María Dolores; Font, Alicia; Garcia-Cano, Jorge

    2015-01-01

    We establish experimentally and through simulations the economic and technical viability of dehydrating ethanol by means of azeotropic distillation, using a hydrocarbon as entrainer. The purpose of this is to manufacture a ready-to-use ethanol–hydrocarbon fuel blend. In order to demonstrate the feasibility of this proposition, we have tested an azeotropic water–ethanol feed mixture, using a hydrocarbon as entrainer, in a semi pilot-plant scale distillation column. Four different hydrocarbons ...

  4. Hydrodeoxygenation of oxidized distilled bio-oil for the production of gasoline fuel type

    International Nuclear Information System (INIS)

    Luo, Yan; Guda, Vamshi Krishna; Hassan, El Barbary; Steele, Philip H.; Mitchell, Brian; Yu, Fei

    2016-01-01

    Highlights: • Oxidation had more influence on the yield of total hydrocarbons than distillation. • The highest total hydrocarbon yield was obtained from oxidized distilled bio-oil. • The 2nd-stage hydrocarbons were in the range of gasoline fuel boiling points. • The main products for upgrading of oxidized bio-oil were aliphatic hydrocarbons. • The main products for upgrading of non-oxidized bio-oil were aromatic hydrocarbons. - Abstract: Distilled and oxidized distilled bio-oils were subjected to 1st-stage mild hydrodeoxygenation and 2nd-stage full hydrodeoxygenation using nickel/silica–alumina catalyst as a means to enhance hydrocarbon yield. Raw bio-oil was treated for hydrodeoxygenation as a control to which to compare study treatments. Following two-stage hydrodeoxygenation, four types of hydrocarbons were mainly comprised of gasoline and had water contents, oxygen contents and total acid numbers of nearly zero and higher heating values of 44–45 MJ/kg. Total hydrocarbon yields for raw bio-oil, oxidized raw bio-oil, distilled bio-oil and oxidized distilled bio-oil were 11.6, 16.2, 12.9 and 20.5 wt.%, respectively. The results indicated that oxidation had the most influence on increasing the yield of gasoline fuel type followed by distillation. Gas chromatography/mass spectrometry characterization showed that 66.0–76.6% of aliphatic hydrocarbons and 19.5–31.6% of aromatic hydrocarbons were the main products for oxidized bio-oils while 35.5–38.7% of aliphatic hydrocarbons and 58.2–63.1% of aromatic hydrocarbons were the main products for non-oxidized bio-oils. Both aliphatic and aromatic hydrocarbons are important components for liquid transportation fuels and chemical products.

  5. 40 CFR 80.620 - What are the additional requirements for diesel fuel or distillates produced by foreign...

    Science.gov (United States)

    2010-07-01

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

  6. Reoxidation of uranium in electrolytically reduced simulated oxide fuel during residual salt distillation

    International Nuclear Information System (INIS)

    Eun-Young Choi; Jin-Mok Hur; Min Ku Jeon; University of Science and Technology, Yuseong-gu, Daejeon

    2017-01-01

    We report that residual salt removal by high-temperature distillation causes partial reoxidation of uranium metal to uranium oxide in electrolytically reduced simulated oxide fuel. Specifically, the content of uranium metal in the above product decreases with increasing distillation temperatures, which can be attributed to reoxidation by Li 2 O contained in residual salt (LiCl). Additionally, we estimate the fractions of Li 2 O reacted with uranium metal under these conditions, showing that they decrease with decreasing temperature, and calculate some thermodynamic parameters of the above reoxidation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-11-01

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

  8. Recovery of Navy distillate fuel from reclaimed product. Volume I. Technical discussion

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-11-01

    In an effort to assist the Navy to better utilize its waste hydrocarbons, NIPER, with support from the US Department of Energy, is conducting research designed to ultimately develop a practical technique for converting Reclaimed Product (RP) into specification Naval Distillate Fuel (F-76). The first phase of the project was focused on reviewing the literature and available information from equipment manufacturers. The literature survey has been carefully culled for methodology applicable to the conversion of RP into diesel fuel suitable for Navy use. Based upon the results of this study, a second phase has been developed and outlined in which experiments will be performed to determine the most practical recycling technologies. It is realized that the final selection of one particular technology may be site-specific due to vast differences in RP volume and available facilities. A final phase, if funded, would involve full-scale testing of one of the recommended techniques at a refueling depot. The Phase I investigations are published in two volumes. Volume 1, Technical Discussion, includes the narrative and Appendices I and II. Appendix III, a detailed Literature Review, includes both a narrative portion and an annotated bibliography containing about 800 referenvces and abstracts. This appendix, because of its volume, has been published separately as Volume 2. 18 figures, 4 tables.

  9. Safe, acceptable anti-microbial strategies for distillate fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hill, E.C. [ECHA Microbiology Ltd., Cardiff (United Kingdom)

    1995-05-01

    Microbiological fouling, spoilage and corrosion have for years been considered as end-user problems but they have now become endemic up-stream affecting cargoes, tank farms and terminals. Trading agreements to share storage and distribution facilities impose the need to mutually agree antimicrobial strategies which satisfy all health, safety and environmental regulations wherever that fuel is distributed and used. Also agreed must be the infection levels at which antimicrobial action is initiated. Physical decontamination methods are described and the use of biocides discussed in relation to increasing regulatory restrictions.

  10. Impact of 50% Synthesized Iso-Paraffins (SIP) on Middle Distillate Fuel Filtration and Coalescence

    Science.gov (United States)

    2014-10-30

    Paraffins DEFINITIONS Coalescence - the ability to shed water Conventional Material Source - crude oil , natural gas liquid condensates...Impact of 50% Synthesized Iso-Paraffins (SIP) on Middle Distillate Fuel Filtration and Coalescence NF&LCFT REPORT 441/15-003 30 October 2014...heavy oil , shale oil , and oil sands Effluent - stream leaving a system Influent - stream entering a system Turnover - time required to flow the

  11. Producing fuel alcohol by extractive distillation: Simulating the process with glycerol

    OpenAIRE

    Ana María Uyazán; Iván Dario Gil; Jaime Aguilar; Gerardo Rodríguez Niño; Luis A Caicedo Mesa

    2006-01-01

    Downstream separation processes in biotechnology form part of the stages having most impact on a product’s final cost. The tendency throughout the world today is to replace fossil fuels with those having a renewable origin such as ethanol; this, in turn, produces a demand for the same and the need for optimising fermentation, treating vinazas and dehydration processes. The present work approaches the problem of dehydration through simulating azeotropic ethanol extractive distillation using gl...

  12. Analysis and Environmental Fate of Air Force Distillate and High Density Fuels

    Science.gov (United States)

    1981-10-01

    728.1 128 0.8 Toluenc 751.3 92 0.6 XTHDCPD 1049.6 136 66.8 NTHDCPD 1079.2 136 1.5HNN 1509.6 186 20.1 JP-1O XTHDCPD 1050.3 136 96.8 ITHDCPD 1079.6 136 1,5...deionized water and the salts listed below. Blanks of both waters were routinely extracted and analyzed for possible 4.nterferences. MNN PXTX XTHDCPD ...through 13; complete data summaries for the distillate fuels may be found in Appendix C. All com- ponents of the high density fuels except XTHDCPD of

  13. Catalytic Hydrotreatment of Light Distillates Obtained from Bio-Oil for Use in Oxygenated Liquid Fuels

    Directory of Open Access Journals (Sweden)

    Xianwei Zheng

    2015-06-01

    Full Text Available Bio-oil can be fractionated into three parts according to their boiling points. This research reported that light distillates could be converted into oxygenated liquid fuels through a two-stage hydrotreatment approach, using a Pd/C catalyst. The main goal of the first hydrotreatment stage was to stabilize the high active components, which contained carbon–carbon double bonds and aldehyde groups. The second hydrotreatment stage aimed to saturate the components with benzene rings and keto groups, resulting in saturated oxygenated compounds. The H/Ceff ratio was improved greatly after the two-stage hydrotreatment, increasing from 0.51 (in the reactant to 1.67 (in the final products. The high heating value of the final products was 31.63 MJ/kg. After the two-stage hydrotreatment, the main products were C5–C9 alcohols, which were tested via gas chromatography–mass spectrometry. The products could be blended with gasoline directly. Based on the experiments regarding the hydrogenated model compounds, a reaction schematic for the two-stage hydrotreatment was created. Moreover, the bio-oil hydrotreatment kinetics were investigated. The order of the hydrotreatment reaction was 2.0, and the apparent activation energy (Ea was 57.29 KJ/mol.

  14. Producing fuel alcohol by extractive distillation: Simulating the process with glycerol

    Directory of Open Access Journals (Sweden)

    Ana María Uyazán

    2006-01-01

    Full Text Available Downstream separation processes in biotechnology form part of the stages having most impact on a product’s final cost. The tendency throughout the world today is to replace fossil fuels with those having a renewable origin such as ethanol; this, in turn, produces a demand for the same and the need for optimising fermentation, treating vinazas and dehydration processes. The present work approaches the problem of dehydration through simulating azeotropic ethanol extractive distillation using glycerol as separation agent. Simulations were done on an Aspen Plus process simulator (Aspen Tech version 11.1. The simulated process involves two distillation columns, a dehydrator and a glycerol recuperation column. Simulation restrictions were ethanol’s molar composition in dehydrator column distillate and the process’s energy consumption. The effect of molar reflux ratio, solvent-feed ratio, solvent entry and feed stage and solvent entry temperature were evaluated on the chosen restrictions. The results showed that the ethanol-water mixture dehydration with glycerol as separation agent is efficient from the energy point of view.

  15. Production of bio-fuel ethanol from distilled grain waste eluted from Chinese spirit making process.

    Science.gov (United States)

    Tan, Li; Sun, Zhaoyong; Zhang, Wenxue; Tang, Yueqin; Morimura, Shigeru; Kida, Kenji

    2014-10-01

    Distilled grain waste eluted from Chinese spirit making is rich in carbohydrates, and could potentially serve as feedstock for the production of bio-fuel ethanol. Our study evaluated two types of saccharification methods that convert distilled grain waste to monosaccharides: enzymatic saccharification and concentrated H2SO4 saccharification. Results showed that enzymatic saccharification performed unsatisfactorily because of inefficient removal of lignin during pretreatment. Concentrated H2SO4 saccharification led to a total sugar recovery efficiency of 79.0 %, and to considerably higher sugar concentrations than enzymatic saccharification. The process of ethanol production from distilled grain waste based on concentrated H2SO4 saccharification was then studied. The process mainly consisted of concentrated H2SO4 saccharification, solid-liquid separation, decoloration, sugar-acid separation, oligosaccharide hydrolysis, and continuous ethanol fermentation. An improved simulated moving bed system was employed to separate sugars from acid after concentrated H2SO4 saccharification, by which 95.8 % of glucose and 85.8 % of xylose went into the sugar-rich fraction, while 83.3 % of H2SO4 went into the acid-rich fraction. A flocculating yeast strain, Saccharomyces cerevisiae KF-7, was used for continuous ethanol fermentation, which produced an ethanol yield of 91.9-98.9 %, based on glucose concentration.

  16. Particle fuel bed tests

    International Nuclear Information System (INIS)

    Horn, F.L.; Powell, J.R.; Savino, J.M.

    1985-01-01

    Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H 2 for 12 hours with no visible reaction or weight loss

  17. Design of a thermally integrated bioethanol-fueled solid oxide fuel cell system integrated with a distillation column

    Science.gov (United States)

    Jamsak, W.; Douglas, P. L.; Croiset, E.; Suwanwarangkul, R.; Laosiripojana, N.; Charojrochkul, S.; Assabumrungrat, S.

    Solid oxide fuel cell systems integrated with a distillation column (SOFC-DIS) have been investigated in this study. The MER (maximum energy recovery) network for SOFC-DIS system under the base conditions (C EtOH = 25%, EtOH recovery = 80%, V = 0.7 V, fuel utilization = 80%, T SOFC = 1200 K) yields Q Cmin = 73.4 and Q Hmin = 0 kW. To enhance the performance of SOFC-DIS, utilization of internal useful heat sources from within the system (e.g. condenser duty and hot water from the bottom of the distillation column) and a cathode recirculation have been considered in this study. The utilization of condenser duty for preheating the incoming bioethanol and cathode recirculation for SOFC-DIS system were chosen and implemented to the SOFC-DIS (CondBio-CathRec). Different MER designs were investigated. The obtained MER network of CondBio-CathRec configuration shows the lower minimum cold utility (Q Cmin) of 55.9 kW and total cost index than that of the base case. A heat exchanger loop and utility path were also investigated. It was found that eliminate the high temperature distillate heat exchanger can lower the total cost index. The recommended network is that the hot effluent gas is heat exchanged with the anode heat exchanger, the external reformer, the air heat exchanger, the distillate heat exchanger and the reboiler, respectively. The corresponding performances of this design are 40.8%, 54.3%, 0.221 W cm -2 for overall electrical efficiency, Combine Heat and Power (CHP) efficiency and power density, respectively. The effect of operating conditions on composite curves on the design of heat exchanger network was investigated. The obtained composite curves can be divided into two groups: the threshold case and the pinch case. It was found that the pinch case which T SOFC = 1173 K yields higher total cost index than the CondBio-CathRec at the base conditions. It was also found that the pinch case can become a threshold case by adjusting split fraction or operating at

  18. Cryogenic distillation: a fuel enrichment system for near-term tokamak-type D-T fusion reactors

    International Nuclear Information System (INIS)

    Misra, B.; Davis, J.F.

    1980-02-01

    The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio

  19. Development and pilot testing of full-scale membrane distillation modules for deployment of waste heat

    NARCIS (Netherlands)

    Jansen, A.E.; Assink, J.W.; Hanemaaijer, J.H.; Medevoort, J. van; Sonsbeek, E. van

    2013-01-01

    Membrane distillation is an attractive technology for extracting fresh water from seawater. Newly developed modules have been used in pilot tests and bench scale tests to demonstrate the potential of producing excellent product water quality in a single step, little need for water pretreatment and a

  20. Development of a test method for distillation of diesel-biodiesel-alcohols mixtures at reduced pressure

    Science.gov (United States)

    Niculescu, R.; Iosub, I.; Clenci, A.; Zaharia, C.; Iorga-Simăn, V.

    2017-10-01

    Increased environmental awareness and depletion of fossil petroleum resources are driving the automotive industry to seek out and use alternative fuels. For instance, the biofuel is a major renewable energy source to supplement declining fossil fuel resources. The addition of alcohols like methanol and ethanol is practical in biodiesel blends due to its miscibility with the pure biodiesel. Alcohols also improve physico-chemical properties of biodiesel blends, which lead to improved combustion efficiency. Proper volatility of fuels is critical to the operation of internal combustion engines with respect to both performance and emissions. Volatility may be characterised by various measurements, the most common of which are vapour pressure, distillation and the vapour/liquid ratio. The presence of ethanol or other oxygenates may affect these properties and, as a result, performance and emissions, as well. However, in the case of diesel-biodiesel-alcohols mixtures, the variance of component volatility makes difficult the analysis of the overall volatility. Thus, the paper presents an experimental method of distilling diesel-biodiesel-alcohols mixtures by adjusting the boiler pressure of an i-Fischer Dist equipment.

  1. Gas Test Loop Booster Fuel Hydraulic Testing

    International Nuclear Information System (INIS)

    Gas Test Loop Hydraulic Testing Staff

    2006-01-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3

  2. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  3. High performance liquid chromatographic hydrocarbon group-type analyses of mid-distillates employing fuel-derived fractions as standards

    Science.gov (United States)

    Seng, G. T.; Otterson, D. A.

    1983-01-01

    Two high performance liquid chromatographic (HPLC) methods have been developed for the determination of saturates, olefins and aromatics in petroleum and shale derived mid-distillate fuels. In one method the fuel to be analyzed is reacted with sulfuric acid, to remove a substantial portion of the aromatics, which provides a reacted fuel fraction for use in group type quantitation. The second involves the removal of a substantial portion of the saturates fraction from the HPLC system to permit the determination of olefin concentrations as low as 0.3 volume percent, and to improve the accuracy and precision of olefins determinations. Each method was evaluated using model compound mixtures and real fuel samples.

  4. Fuel Economy Testing and Data

    Science.gov (United States)

    EPA’s Fuel Economy pages provide information on current standards and how federal agencies work to enforce those laws, testing for national Corporate Average Fuel Economy or CAFE standards, and what you can do to reduce your own vehicle emissions.

  5. Worldwide Distillate Fuel Quality and Engine Technology Through the Year 2010

    National Research Council Canada - National Science Library

    Westbrook, S

    2000-01-01

    This report gives the results of a program to assess the potential changes in future petroleum refining technology and the potential changes in the properties of future commercial, distillate, marine...

  6. ITER task T48 (1994); low-inventory cryogenic distillation tests

    International Nuclear Information System (INIS)

    Woodall, K.; Robins, J.; Bellamy, D.

    1995-01-01

    Previous work at Ontario Hydro Technologies (OHT) had shown that small cryogenic columns could be stably controlled and designed to much lower inventories than had been previously thought possible. Among the results were measurements of Height-of-Equivalent-Theoretical-Plate (HETP) versus holdup for Heli-Pak A and B in columns up to 20 mm diameter. ITER cryogenic distillation column designs suggest that the final high-tritium columns could be 30-70 mm diameter. The objective of this ITER task was to design and construct a column section for demonstration of scale-up of low inventory cryogenic distillation. The experiments were to be carried out in an upgraded Cryogenics Distillation Laboratory at OHT, in the facility used for previous low-inventory column tests. The ITER scaled-up test system as the following characteristics: 55 W condenser capacity; 30 mm diameter column loaded with Helipak B; 1500 mm packed height. The first task was to design and build the scaled-up test facility. In order to reduce costs, it was necessary to use existing 30-35 W helium refrigerators. Therefore, to provide 60-W duty to the scaled-up column, the two refrigerators had to be well coupled thermally, but not mechanically, since the refrigerator cold heads have very thin shells. The solution was to attach the column firmly to one cold head and indirectly to an adjacent cold head through flexible copper braid. Several iterations were required to obtain the desired good heat transfer with flexible mechanical connection. This facility is now operational and ready to begin measurements on the 30 mm column. Also during 1994, the Princeton Tritium Processing System (TPS) was installed and commissioned. The results from this experience are relevant to the ITER distillation system. 2 refs., 10 figs

  7. ITER task T48 (1994); low-inventory cryogenic distillation tests

    Energy Technology Data Exchange (ETDEWEB)

    Woodall, K; Robins, J; Bellamy, D [Ontario Hydro, Toronto, ON (Canada). Research Div.; Sood, S; Fong, C [Ontario Hydro, Toronto, ON (Canada)

    1995-01-01

    Previous work at Ontario Hydro Technologies (OHT) had shown that small cryogenic columns could be stably controlled and designed to much lower inventories than had been previously thought possible. Among the results were measurements of Height-of-Equivalent-Theoretical-Plate (HETP) versus holdup for Heli-Pak A and B in columns up to 20 mm diameter. ITER cryogenic distillation column designs suggest that the final high-tritium columns could be 30-70 mm diameter. The objective of this ITER task was to design and construct a column section for demonstration of scale-up of low inventory cryogenic distillation. The experiments were to be carried out in an upgraded Cryogenics Distillation Laboratory at OHT, in the facility used for previous low-inventory column tests. The ITER scaled-up test system as the following characteristics: 55 W condenser capacity; 30 mm diameter column loaded with Helipak B; 1500 mm packed height. The first task was to design and build the scaled-up test facility. In order to reduce costs, it was necessary to use existing 30-35 W helium refrigerators. Therefore, to provide 60-W duty to the scaled-up column, the two refrigerators had to be well coupled thermally, but not mechanically, since the refrigerator cold heads have very thin shells. The solution was to attach the column firmly to one cold head and indirectly to an adjacent cold head through flexible copper braid. Several iterations were required to obtain the desired good heat transfer with flexible mechanical connection. This facility is now operational and ready to begin measurements on the 30 mm column. Also during 1994, the Princeton Tritium Processing System (TPS) was installed and commissioned. The results from this experience are relevant to the ITER distillation system. 2 refs., 10 figs.

  8. Design and test of a vacuum distillation method for determining carbon in sodium

    International Nuclear Information System (INIS)

    Irmisch, R.; Rettig, D.; Woelke, K.

    1976-08-01

    A method is described for determining total and carbonate carbon in sodium samples until 10 g. Sodium is removed by vacuum distillation at 300 0 C and the carbon in the residue is converted to carbon dioxide by combustion in a stream of air or thermic splitting in a stream of cover gas at 1000 0 C. The carbon dioxide is measured manometrically. It is therefore not necessary to carry out calibration. Distillation and combustion rig are combined with inertgas filled transfer box. Therefore the sodium sample does not get into touch with air. Test of this method was carried out with Na 2 CO 3 and WC. Carbon recoveries were for Na 2 CO 3 between 103 and 107% and for WC between 92 and 96%. The blank value found being 9 μg C and sensitivity 3 μg C. (author)

  9. ITER task T332a (1996) low-inventory cryogenic distillation tests

    International Nuclear Information System (INIS)

    Woodall, K.; Bellamy, D.

    1997-02-01

    The overall objective of this ITER task was to perform tests which would improve the accuracy of the ITER Isotope Separation System (ISS) tritium inventory estimates and to allow designers to lower the tritium and hydrogen inventory estimates. The work program was also designed to give a better understanding of cryogenic distillation hydraulics and provide information which would improve process control. The work program this year addressed the following specific task objectives. Measure the detailed hydraulics for deuterium/deuterium hydride mixtures in a cryogenic distillation column using Helipak C packing. Determine maximum vapour velocity, HETP and tritium and deuterium inventory data for a column that can be operated right up to flooding conditions. Determine if a proprietary surface treatment improves the wetting characteristics for hydrogen on stainless steel packing. Measure the isotope separation and inventory performance of a Helipak C column large enough to handle up to 45 mm. Investigate additional hydraulic effects in the reboiler and column. 7 refs., 7 figs

  10. Waste acid detoxification and reclamation: Final Phase II report: Summary of distillation and material degradation tests for FY 1987 and FY 1988

    International Nuclear Information System (INIS)

    Brouns, T.M.; Stewart, T.L.

    1988-11-01

    The objective of the Waste Acid Detoxification and Reclamation project is to develop processes for reducing the volume, quantity, and toxicity of metal-bearing waste acids. The primary incentives for implementing these types of waste minimization processes are regulatory and economic (that is, to meet requirements in the Resource Conservation and Recovery Act and reduce the cost for treatment, storage, and disposal). Two precipitation processes and a distillation process are being developed to minimize waste from fuel fabrication operations or other metal-finishing operations conducted at US Department of Energy Defense Programs (DOE-DP) facilities. Waste process acids such as (1) HF-HNO 3 etch solutions containing Zr as a major metal impurity, and (2) HNO 3 strip solutions containing Cu as a major metal impurity are detoxified and reclaimed by concurrently precipitating heavy metals and regenerating acid for recycle. Acid from a third waste acid stream generated from chemical milling operations is reclaimed by the use of distillation. This stream comprises HNO 3 and H 2 SO 4 containing U as the major metal impurity. Distillation allows NO 3 /sup /minus// to be displaced by SO 4 /sup /minus/2/ in metal salts; free HNO 3 is then vaporized from the U-bearing sulfate stream. In a downstream precipitation step, uranium can be recovered from the sulfate stream. This report summarizes distillation tests and candidate material degradation tests conducted during FY 1987 and FY 1988

  11. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Bennie, G

    1875-11-12

    For distilling shale, or other oil-yielding minerals, two or more, preferably four, vertical retorts are mounted in a brickwork oven and are heated in the first place by coal, coke or other fuel on a grate. The spent material from the retorts is discharged from one or more in turn on to the grate and is used, together with additional fuel if necessary, to maintain the heat of the retorts. The retorts are charged by means of hoppers and lids and are discharged by means of movable bottoms actuated by rods and levers acting in combination with outlet valves. The retorts are tapered from the bottom upwards.

  12. Results and Interpretation of the WFRD ELS Distillation Down-Select Test Data

    Science.gov (United States)

    Delzeit, Lance Dean; Flynn, Michael; Carter, Layne; Long, David A.

    2010-01-01

    Testing of the Wiped-film Rotating-disk (WFRD) evaporator was conducted in support of the Exploration Life Support Distillation Down-Select Test. The WFRD was constructed at NASA Ames Research Center (ARC) and tested at NASA Marshall Space Flight Center (MSFC). The WFRD was delivered to MSFC in September 2009, and testing of solution #1 and solution #2 immediately following. Solution #1 was composed of humidity condensate and urine, including flush water and pretreatment chemicals. Solution #2 was composed of hygiene water, humidity condensate, and urine, including flush water and pretreatment chemicals. During the testing, the operational parameters of the WFRD were recorded and samples of the feed, brine, and product were collected and analyzed. The steady-state results of processing 414L of feed solution #1 and 1283L of feed solution #2 demonstrated that running the WFRD at a brine temperature of 50 C gave an average production rate of 16.7 L/hr. The specific energy consumption was 80.5W-hr/L. Data Analysis shows that the water recovery rates were 94% and 91%, respectively. The total mass of the WFRD as delivered to MSFC was 300 Kg. The volume of the tests stand rack was 1m width x 0.7m depth x 1.9m height or 1.5 cu m of which about half of the total volume is occupied by equipment. Chemical analysis of the distillate showed an average TOC of 20ppm, a pH of 3.5, and a conductivity of 98 mho/cm. The conductivity of the distillate, compared to the feed, decreased by 98.9%., the total ion concentration decreased by 99.6%, the total organics decreased 98.6%, and the metals were at or below detection limits

  13. Performance test of solar energy distillation system; Taiyonetsu riyo kaisui tansuika system no zosui seino shiken

    Energy Technology Data Exchange (ETDEWEB)

    Kawahara, T; Toyoshima, Y [Keio University, Tokyo (Japan); Narasaki, Y; Kamiya, I [Ebara Corp., Tokyo (Japan); Sato, H [Keio University, Tokyo (Japan). Faculty of Science and Technology

    1997-11-25

    In order to develop a seawater desalination equipment utilizing solar heat, a performance test was performed on a three-stage solar heat vacuum distiller. The equipment can evaporate and condense water at low temperatures by reducing the pressure in a distilling section. Solar heat is collected by a heat collecting plate, by which water as the working fluid is evaporated and moved to a heat conducting section in an evaporation pipe in a state of steam. The steam is condensed in a low-temperature heat conducting pipe, the heat is released outside the heat conducting pipe, and seawater is evaporated under low pressure condition. The working fluid is circulated for repetitive use. Solar cells are used as a motive power source for a vacuum pump, hence the system is self-sustainable even if installed in a desert area and the like places. The construction is simple, highly durable, and easy in maintenance. The system has high water producing performance, uses no harmful substances whatsoever such as fluorocarbons, and is a friendly system to the global environment. Because of evaporation at low temperatures, heat loss to the surroundings and sensible heat due to temperature rise are small, and the system response is quick. The solar heat can be utilized more effectively as multiple effects. When a heat collecting plate of 7.76 m {sup 2} was used, maximum yield of 105.45 kg/day was obtained. The system`s distillation efficiency is higher than other types of solar heat utilizing distillers. 7 refs., 5 figs.

  14. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Cosden, S; Cosden, J S

    1937-09-08

    A means and process are described for the destructive distillation of solid carbonaceous materials in which the process comprises charging the material, in a finely divided condition into a stream of hot combustion gases, and allows the hot gases to act pyrolytically on the organic compounds contained in the material, separating the volatile liberated constituents from residuary constituents. Hot reaction gases are generated by fuel ignition means in a generator and are immediately intermingled with comminuted carbonaceous material from a hopper, in a narrow conduit. The mixture of material and reaction fluid is then passed through an elongated confined path, which is exteriorly heated by the combustion chamber of the furnace, where the destructive distillation is effected. Volatile and solid constituents are separated in the chamber, and the volatile constituents are fractionated and condensed.

  15. Distillers Dried Grains with Solubles (DDGS) – A Key to the Fuel Ethanol Industry

    Science.gov (United States)

    Corn-based ethanol in the U.S. has dramatically increased in recent years; so has the quantity of associated coproducts. Nonfermentable components are removed from the process as whole stillage, centrifuged to remove water – which is then evaporated to produce condensed distillers solubles (CDS), a...

  16. CANFLEX fuel bundle impact test

    International Nuclear Information System (INIS)

    Chang, Seok Kyu; Chung, C. H.; Park, J. S.; Hong, S. D.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs

  17. CANFLEX fuel bundle strength tests (test report)

    International Nuclear Information System (INIS)

    Chang, Seok Kyu; Chung, C. H.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs

  18. Process of coking or distilling fuels, bituminous shale, and the like

    Energy Technology Data Exchange (ETDEWEB)

    Jurgensen, R

    1892-03-17

    The charge present in a chamber is exposed without changes to stepwise higher temperatures. This is accomplished so that only a few chambers whose contents are almost finished are heated by generator gas. The distillation products of these chambers go through the charge of the following chambers and carry out carbonization products, which are given for the most part by passage through the following chambers to the colder charge in condensed form, whereby the gases escape out the chimney.

  19. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Dow, P

    1884-09-11

    The invention has for its object to regulate the temperature at any point of vertical retorts in which steam or steam and air are employed for the distillation of shale, coal, and other substances. Vertical steam pipes at the exterior of each retort and connected with main pipes have a series of branches at different levels and furnished with regulating-valves or cocks. The admission of air is similarly regulated and spy-holes with shutters blocked or sealed against the escape of such products by the fuel intake at one end of the conduit and the congested masses of coke discharged at the other.

  20. Application of the Advanced Distillation Curve Method to Fuels for Advanced Combustion Engine Gasolines

    KAUST Repository

    Burger, Jessica L.; Schneider, Nico; Bruno, Thomas J.

    2015-01-01

    . To make changes in the most time- and cost-effective manner, it is imperative that new computational tools and surrogate fuels are developed. Currently, sets of fuels are being characterized by industry groups, such as the Coordinating Research Council

  1. ITER Task T332a (1995): Low-inventory cryogenic distillation tests

    International Nuclear Information System (INIS)

    Woodall, K.; Robins, J.; Bellamy, D.

    1996-01-01

    Previous work at Ontario Hydro Technologies (OHT) had shown that small hydrogen cryogenic columns could be stably controlled and designed to much lower inventories than had been previously thought possible. Among the results were measurements of height equivalent to a theoretical plate (HETP) versus holdup for Helipak A and B packings in columns up to 20 mm diameter. ITER cryogenic distillation column designs suggest that the final high-tritium column could be 30-70 mm diameter. The objective of this ITER task was to test scale-up of OHT low inventory columns to ITER dimensions. In 1994 OHT built a suitable test facility. In 1995, two low-inventory packings were tested. (author) 4 refs., 6 figs

  2. Application of the Advanced Distillation Curve Method to the Comparison of Diesel Fuel Oxygenates: 2,5,7,10-Tetraoxaundecane (TOU), 2,4,7,9-Tetraoxadecane (TOD), and Ethanol/Fatty Acid Methyl Ester (FAME) Mixtures.

    Science.gov (United States)

    Burger, Jessica L; Lovestead, Tara M; LaFollette, Mark; Bruno, Thomas J

    2017-08-17

    Although they are amongst the most efficient engine types, compression-ignition engines have difficulties achieving acceptable particulate emission and NO x formation. Indeed, catalytic after-treatment of diesel exhaust has become common and current efforts to reformulate diesel fuels have concentrated on the incorporation of oxygenates into the fuel. One of the best ways to characterize changes to a fuel upon the addition of oxygenates is to examine the volatility of the fuel mixture. In this paper, we present the volatility, as measured by the advanced distillation curve method, of a prototype diesel fuel with novel diesel fuel oxygenates: 2,5,7,10-tetraoxaundecane (TOU), 2,4,7,9-tetraoxadecane (TOD), and ethanol/fatty acid methyl ester (FAME) mixtures. We present the results for the initial boiling behavior, the distillation curve temperatures, and track the oxygenates throughout the distillations. These diesel fuel blends have several interesting thermodynamic properties that have not been seen in our previous oxygenate studies. Ethanol reduces the temperatures observed early in the distillation (near ethanol's boiling temperature). After these early distillation points (once the ethanol has distilled out), B100 has the greatest impact on the remaining distillation curve and shifts the curve to higher temperatures than what is seen for diesel fuel/ethanol blends. In fact, for the 15% B100 mixture most of the distillation curve reaches temperatures higher than those seen diesel fuel alone. In addition, blends with TOU and TOD also exhibited uncommon characteristics. These additives are unusual because they distill over most the distillation curve (up to 70%). The effects of this can be seen both in histograms of oxygenate concentration in the distillate cuts and in the distillation curves. Our purpose for studying these oxygenate blends is consistent with our vision for replacing fit-for-purpose properties with fundamental properties to enable the development of

  3. Distilling tar; distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Brash, P; Young, W

    1866-09-17

    The tarry residue, which separates on treating crude shale oil with sulfuric acid, is redistilled, in the manner described in Specification No. 1278, A.D. 1866, together with shale. Previous to the distillation, the acid is neutralized with lime, or may be separated by blowing steam into the tar and adding salt. The purified tar thus obtained is absorbed by ashes, or is mixed with lime or other alkaline matter, or the shale may be mixed with lime and distilled with the tar, which is allowed to flow over and through the shale during the process. The tar obtained in the purification of natural paraffin may be similarly utilized.

  4. Investigations of Trace Oxygenates in Middle Distillate Fuels using Gas Chromatography

    OpenAIRE

    RENEE LOUISE WEBSTER

    2017-01-01

    There can be up to one million different compounds in aviation or diesel fuels, making the analysis of trace components within the complex matrix highly challenging. Many trace oxygenated compounds may be present in fuels and can have dramatic effects on the fuel’s properties. Advanced analytical chemistry techniques have been used to contribute a critical understanding of the role of trace oxygenates on the chemistry of both emerging alternate and fossil fuels. Knowledge of these molecular s...

  5. Distilling shale and the like

    Energy Technology Data Exchange (ETDEWEB)

    Gee, H T.P.

    1922-02-23

    In distilling shale or like bituminous fuels by internal heating with hot gas obtained by the gasifying of the shale residues with air or steam or a mixture of these, the amount and temperature of the gaseous distilling medium is regulated between the gasifying and the distilling chambers, by the introduction of cold gas or air.

  6. Fuel Subsystems Flight Test Handbook

    Science.gov (United States)

    1981-12-01

    detailed, accessible .-ltand complete test records for his own protection and for the benefit of his successor in case of promotion, transfer or...and pilot display of fuel quantity, low level warning and a " Bingo " fudl warning. 3.0 TEST OBJECT1VES: ACTION Orriclc. On POUTION•’PHONWE [ DATE PCR...TIS No. 46, paragraph 3.11 3.6 To demonstrate that the low level and bingo warning system are consistent L •and meet the requirements of paragraph

  7. Verification tests for CANDU advanced fuel

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1997-07-01

    For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs

  8. Leak testing fuel stored in the ICPP fuel storage basin

    International Nuclear Information System (INIS)

    Lee, J.L.; Rhodes, D.W.

    1977-06-01

    Irradiated fuel to be processed at the Idaho Chemical Processing Plant is stored under water at the CPP-603 Fuel Storage Facility. Leakage of radionuclides through breaks in the cladding of some of the stored fuels contaminates the water with radionuclides resulting in radiation exposure to personnel during fuel handling operations and contamination of the shipping casks. A leak test vessel was fabricated to test individual fuel assemblies which were suspected to be leaking. The test equipment and procedures are described. Test results demonstrated that a leaking fuel element could be identified by this method; of the eleven fuel assemblies tested, six were estimated to be releasing greater than 0.5 Ci total radionuclides/day to the basin water

  9. Modeling RP-1 Fuel Advanced Distillation Data using Comprehensive Two-Dimensional Gas Chromatography Coupled with Time-of-Flight Mass Spectrometry and Partial Least Squares Analysis

    Science.gov (United States)

    2014-05-07

    of a variety of complex liquid samples including, but not limited to, crude oil [12], gasoline [16], biodiesel fuel [17, 19], jet fuel [5, 10, 11...measurements for corrosive fluids: Application to two crude oils . Fuel 87:3055–3064. doi: 10.1016/j.fuel.2008.04.032 13. Ott LS, Hadler AB, Bruno TJ...Windom BC, Bruno TJ (2013) Application of Pressure-Controlled Advanced Distillation Curve Analysis: Virgin and Waste Oils . Ind Eng Chem Res 52:327–337

  10. Recent metal fuel safety tests in TREAT

    International Nuclear Information System (INIS)

    Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

    1986-01-01

    In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

  11. Accident analysis of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.; Chi, D. Y

    1998-03-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

  12. Microbial development in distillers wet grains produced during fuel ethanol production from corn (Zea mays)

    Energy Technology Data Exchange (ETDEWEB)

    Lehman, R.M.; Rosentrater, K.A. [United States Dept. of Agriculture, Brookings, SD (United States). North Central Agricultural Research Laboratory

    2007-09-15

    The microbiology of post-production distillers wet grains (DWG) was investigated over a period of 9 days at an industrial ethanol plant. Samples of the DWG were physically and chemically characterized. Compositional analyses were conducted for protein, fiber, and fat. Fixed suspensions of DWG were dispersed and disrupted by sonication. Bacterial cells were enumerated under epifluorescent illumination. Solid media and standard dilution were used to enumerate total colony-forming units (CFU) of lactic-acid producing bacteria (LAB), and aerobic heterotrophic organisms. The DWG had a pH of approximately 4.4, a moisture content of 53.5 per cent, and 4 x 10{sup 5} total yeast cells. Thirteen morphologically distinct isolates were identified during the study, 10 of which were yeasts and molds from 6 different genera. Two of the yeasts were of the lactic-acid Pediococcus pentosaceus strain, and 1 of the yeasts was an aerobic heterotrophic bacteria. Results showed that the matrix of the DWG produced severe technical difficulties for several of the culture-independent community-level analyses. It was concluded that numbers of potentially beneficial bacteria appeared to increase over the time period relative to potential spoilage agents. Molds capable of producing mycotoxins colonized the DWG and grew to high densities over the 9 day period. 31 refs., 3 tabs., 2 figs.

  13. Reference test methods for total water in lint cotton by Karl Fischer Titration and low temperature distillation

    Science.gov (United States)

    In a study of comparability of total water contents (%) of conditioned cottons by Karl Fischer Titration (KFT) and Low Temperature Distillation (LTD) reference methods, we demonstrated a match of averaged results based on a large number of replications and weighing the test specimens at the same tim...

  14. Comparative TEA for Indirect Liquefaction Pathways to Distillate-Range Fuels via Oxygenated Intermediates

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Eric; Snowden-Swan, Lesley J.; Talmadge, Michael; Dutta, Abhijit; Jones, Susanne; Ramasamy, Karthikeyan; Gray, Michael; Dagle, Robert; Padmaperuma, Asanga; Gerber, Mark; Sahir, Asad; Tao, Ling; Zhang, Yanan

    2017-03-03

    This paper presents a comparative techno-economic analysis of five conversion pathways from biomass to gasoline-, jet-, and diesel-range hydrocarbons via indirect liquefaction with specific focus on pathways utilizing oxygenated intermediates (derived either via thermochemical or biochemical conversion steps). The four emerging pathways of interest are compared with one conventional pathway (Fischer-Tropsch) for the production of the hydrocarbon blendstocks. The processing steps of the four emerging pathways include: biomass-to-syngas via indirect gasification, gas cleanup, conversion of syngas to alcohols/oxygenates, followed by conversion of alcohols/oxygenates to hydrocarbon blendstocks via dehydration, oligomerization, and hydrogenation. We show that the emerging pathways via oxygenated intermediates have the potential to be cost competitive with the conventional Fischer-Tropsch process. The evaluated pathways and the benchmark process generally exhibit similar fuel yields and carbon conversion efficiencies. The resulting minimum fuel selling prices are comparable to the benchmark at approximately $3.60 per gallon-gasoline equivalent, with potential for two new pathways to be more economically competitive. Additionally, the coproduct values can play an important role in the economics of the processes with oxygenated intermediates derived via syngas fermentation. Major cost drivers for the integrated processes are tied to achievable fuel yields and conversion efficiency of the intermediate steps, i.e., the production of oxygenates/alcohols from syngas and the conversion of oxygenates/alcohols to hydrocarbon fuels.

  15. DESIGN, DEVELOPMENT AND TESTING OF MULTIEFFECT DISTILLER/EVAPORATOR USING SCHEFFLER SOLAR CONCENTRATORS

    Directory of Open Access Journals (Sweden)

    AJAY CHANDAK

    2009-09-01

    Full Text Available Authors designed and experimented with multistage evaporation system for production of distilled water. Two Scheffler concentrators of 16 sqm each were used for generating steam in the first stage at 8 bar pressure and the pressure is gradually brought down to 1 bar, in four stage distillation unit. Total yield obtained in the project was 2.3 times that of single stage distillation. Temperature drop in every subsequent stage was designed to 25 degree centigrade. Heat of condensation in the last stage and also sensible heat of the condensate in all the stages were used for preheating of water in the next batch. The system has great potential in food processing industry for applications of juice thickening, sauces, jams, salt concentrating systems and distilled water applications, desalination etc. Results of the project are very encouraging for commercial scale application.

  16. Technical specification of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y

    1998-03-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the operation limit, safety limit, operation condition and checking points of HANARO fuel test loop. This results will become guidances for the planning of irradiation testing and operation of HANARO fuel test loop. (author). 13 refs., 13 tabs., 8 figs.

  17. Sodium distiller II

    International Nuclear Information System (INIS)

    Goncalves, A.C.; Castro, P.M. e; Torres, A.R.; Correa, S.M.

    1990-01-01

    A sodium distiller allows the evaluation of the sodium purity, contained in plants and circuits of Fast Reactors. The sodium distillers of the IEN Reactor's Department was developed initially as a prototype, for the testing of the distillation process and in a second step, as a equipment dedicated to attendance the operation of these circuits. This last one was build in stainless steel, with external heat, rotating crucible of nickel for four samples, purge system for pipe cleaning and a sight glass that permits the observation of the distillation during all the operation. The major advantage of this equipment is the short time to do a distillation operation, which permits its routine utilization. As a consequence of the development of the distillers and its auxiliary systems an important amount of new information was gathered concerning components and systems behaviour under high temperature, vacuum and sodium. (author)

  18. Distillation methods

    International Nuclear Information System (INIS)

    Konecny, C.

    1975-01-01

    Two main methods of separation using the distillation method are given and evaluated, namely evaporation and distillation in carrier gas flow. Two basic apparatus are described for illustrating the methods used. The use of the distillation method in radiochemistry is documented by a number of examples of the separation of elements in elemental state, volatile halogenides and oxides. Tables give a survey of distillation methods used for the separation of the individual elements and give conditions under which this separation takes place. The suitability of the use of distillation methods in radiochemistry is discussed with regard to other separation methods. (L.K.)

  19. Energy deposition in NSRR test fuels

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Tanzawa, Sadamitsu; Tanzawa, Tomio; Kitano, Teruaki; Okazaki, Shuji

    1978-02-01

    Interpretation of fuel performance data collected during inpile testing in the NSRR requires a knowledge of the energy deposition or enthalpy increase in each sample tested. The report describes the results of absolute measurement of fission products and contents of uranium in irradiated test fuels which were performed to determine the energy deposition. (auth.)

  20. Fuel Cell Stations Automate Processes, Catalyst Testing

    Science.gov (United States)

    2010-01-01

    Glenn Research Center looks for ways to improve fuel cells, which are an important source of power for space missions, as well as the equipment used to test fuel cells. With Small Business Innovation Research (SBIR) awards from Glenn, Lynntech Inc., of College Station, Texas, addressed a major limitation of fuel cell testing equipment. Five years later, the company obtained a patent and provided the equipment to the commercial world. Now offered through TesSol Inc., of Battle Ground, Washington, the technology is used for fuel cell work, catalyst testing, sensor testing, gas blending, and other applications. It can be found at universities, national laboratories, and businesses around the world.

  1. Conventional fuel tank blunt impact tests : test and analysis results

    Science.gov (United States)

    2014-04-02

    The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. A series of impact tests are planned to : measure fuel tank deformation under two types of dynamic : loading conditi...

  2. Test requirements of locomotive fuel tank blunt impact tests

    Science.gov (United States)

    2013-10-15

    The Federal Railroad Administrations Office of Research : and Development is conducting research into passenger : locomotive fuel tank crashworthiness. A series of impact tests : are planned to measure fuel tank deformation under two types : of dy...

  3. HANARO fuel irradiation test (II): revision

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

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

  4. Dry distillation

    Energy Technology Data Exchange (ETDEWEB)

    1939-11-30

    To heat rapidly, and prevent agglutination of carbonaceous material duriing distillation of shale, a furnace of the tunnel type has four compartments (the preheating chamber, the distillation chamber proper, and two cooling chambers). Waggons, which convey the material through the distilling chamber, have perforated bottoms. Above the waggons in the distilling chamber are three heating sections having pipes which pass through the sections and communicate with the distilling chamber. Fans cause the distillation gases to circulate through the material and the pipes. The heating gases from three fire boxes are introduced into the oven, and circulate around pipes and are drawn to the discharge apertures by the fans. The heating gases introduced at two points travel in the direction of the material being treated, while the gases introduced at a third point travel in counter flow thereto. Gas is discharged by two pipes. Trucks carrying treated material are passed to two cooling chambers.

  5. Spent fuel's behavior under dynamic drip tests

    International Nuclear Information System (INIS)

    Finn, P.A.; Buck, E.C.; Hoh, J.C.; Bates, J.K.

    1995-01-01

    In the potential repository at Yucca Mountain, failure of the waste package container and the cladding of the spent nuclear fuel would expose the fuel to water under oxidizing conditions. To simulate the release behavior of radionuclides from spent fuel, dynamic drip and vapor tests with spent nuclear fuel have been ongoing for 2.5 years. Rapid alteration of the spent fuel has been noted with concurrent release of radionuclides. Colloidal species containing americium and plutonium have been found in the leachate. This observation suggests that colloidal transport of radionuclides should be included in the performance assessment of a potential repository

  6. Shale distillation

    Energy Technology Data Exchange (ETDEWEB)

    Blanding, F H

    1946-08-29

    A continuous method of distilling shale to produce valuable hydrocarbon oils is described which comprises providing a fluidized mass of the shale in a distillation zone, withdrawing hydrocarbon vapors from the zone, mixing fresh cold shale with the hydrocarbon vapors to quench the same, whereby the fresh shale is preheated, recovering hydrocarbon vapors and product vapors from the mixture and withdrawing preheated shale from the mixture and charging it to a shale distillation zone.

  7. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    1938-07-05

    A process and apparatus for the destructive distillation at low temperature of mineral or organic material particularly oil shale, is given in which the process comprises distilling the material in a horizontal gaseous stream, subjecting the hot residues to the action of a gaseous stream containing a predetermined amount of oxygen so as to burn, at least partly, the carbon-containing substances, and the process uses the gases from this combustion for the indirect heating of the gases serving for the distillation.

  8. Distilling shale

    Energy Technology Data Exchange (ETDEWEB)

    Heyl, G E

    1917-02-06

    The yield of oil obtained by distilling shale is increased by first soaking the shale with about 10 percent of its volume of a liquid hydrocarbon for a period of 24 hours or longer. Distillation is carried on up to a temperature of about 220/sup 0/C., and a further 10 percent of hydrocarbon is then added and the distillation continued up to a temperature of about 400/sup 0/C.

  9. HANARO fuel irradiation test(II)

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

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

  10. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  11. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  12. FCTESTNET - Testing fuel cells for transportation

    NARCIS (Netherlands)

    Winkel, R.G.; Foster, D.L.; Smokers, R.T.M.

    2006-01-01

    FCTESTNET (Fuel Cell Testing and Standardization Network) is an ongoing European network project within Framework Program 5. It is a three-year project that commenced January 2003, with 55 partners from European research centers, universities, and industry, working in the field of fuel cell R and D.

  13. Status of fuel irradiation tests in HANARO

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

    1999-01-01

    Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

  14. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  15. Low temperature distillation of coal, shale, etc

    Energy Technology Data Exchange (ETDEWEB)

    1938-08-12

    A process is disclosed for the low temperature distillation of solid carbonaceous fuels, such as coal, lignite, shale or the like, which comprises feeding or supplying the comminuted fuel in the form of a layer of shallow depth to drying and distilling zones in succession moving the fuel forward through the zones, submitting it to progressively increasing nonuniform heating therein by combustion gases supplied to the distillation zone and traveling thence to the drying zone, the gases heating the distillation zone indirectly and the drying zone both indirectly and then directly such that the fuel retains its solid discrete form during substantially the whole of its travel through the drying and distillation zones, subjecting the fuel for a portion of its travel to a zigzag ploughing and propelling movement on a heated sole, and increasing the heating so as to cause fusion of the fuel immediately prior to its discharge from the distillation zone.

  16. Development of an Exploration-Class Cascade Distillation Subsystem: Performance Testing of the Generation 1.0 Prototype

    Science.gov (United States)

    Callahan, Michael R.; Sargusingh, Miriam J.

    2015-01-01

    The ability to recover and purify water is crucial for realizing long-term human space missions. The National Aeronautics and Space Admininstration and Honeywell co-developed a five-stage vacuum rotary distillation water recovery system referred to as the Cascade Distillation Subsystem (CDS). Over the past three years, NASA's Advanced Exploration Systems (AES) Water Recovery Project (WRP) has been working toward the development of a flight-forward CDS design. In 2012 the original CDS prototype underwent a series of incremental upgrades and tests intened to both demonstrate the feasibility of a on-orbit demonstration of the system and to collect operational and performance data to be used to inform a second generation design. The latest testing of the CDS Generation 1.0 prototype was conducted May 29 through July 2, 2014. Initial system performance was benchmarked by processing deionized water and sodium chloride. Following, the system was challenged with analogue urine waste stream solutions stabilized with an Oxone-based and the two International Space Station baseline and alternative pretreatment solutions. During testing, the system processed more than 160 kilograms of wastewater with targeted water recoveries between 75 and 85% depending on the specific waste stream tested. For all wastewater streams, contaminant removals from wastewater feed to product water distillate, were estimated at greater than 99%. The average specific energy of the system was less than 120 Watt-hours/kilogram. The following paper provides detailed information and data on the performance of the CDS as challenged per the WRP test objectives.

  17. Future Transient Testing of Advanced Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  18. Future Transient Testing of Advanced Fuels

    International Nuclear Information System (INIS)

    Carmack, Jon

    2009-01-01

    The transient in-reactor fuels testing workshop was held on May 4-5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat energie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric - Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the

  19. RIA tests in CABRI with MOX fuel

    International Nuclear Information System (INIS)

    Schmitz, F.; Papin, J.; Gonnier, C.

    2000-01-01

    Three MOX-fuel tests have been successfully performed within the framework of the CABRI REP-Na test program. From the experimental findings which are presently available, no evidence for thermal effects resulting from the heterogeneous nature of the fuel can be given. There are very clear hints however that fission gas effects are enhanced with regard to the behaviour of UO 2 . The clad rupture observed in REP-Na 7 is of different nature than the failures observed in Cabri tests with UO 2 fuel. Failures of UO 2 fuel rods only occurred when the clad mechanical properties were severely affected by the presence of hydride blisters, while in REP-Na 7 a clear indication is made that the loading potential of the MOX fuel pellets was high enough to break a sound cladding. Concerning the transient fuel behaviour after reaching the critical heat-flux under reactor typical conditions (pressure, temperature and flow), no data base could be provided by the tests in the present sodium test loop (as for the UO 2 fuel behaviour). The IPSN project to implement into the Cabri reactor a pressurised water loop which will allow to simulate the complete RIA accident sequence under PWR reactor typical conditions, aims at providing this missing data base. (author)

  20. MTR fuel testing in BR2

    International Nuclear Information System (INIS)

    Jacquet, P.; Verwimp, A.; Wirix, S.

    2000-01-01

    New fuel design for MTR 's requires to be qualified under representative conditions, that is geometry, neutron spectrum, heat flux and thermo hydraulic conditions. An irradiation device for fuel plates has been designed to derive the maximum benefit from the BR2 irradiation capacities. The fuel plates can be easily extracted from their support during a shutdown to undergo additional tests. One of these tests is the measurement of the thickness changes along the fuel plate. To that purpose, a facility in the reactor water pool has been designed to measure the fuel swelling with an accuracy of 5 μm using inductive probes. At SCK-CEN, the full range of destructive and non-destructive PIE can be performed, including γ-scanning, wet sipping, surface examination and other methods. (author)

  1. Test report : alternative fuels propulsion durability evaluation

    Science.gov (United States)

    2012-08-28

    This document, prepared by Honeywell Aerospace, Phoenix, AZ (Honeywell), contains the final : test report (public version) for the U.S. Department of Transportation/Federal Aviation : Administration (USDOT/FAA) Alternative Fuels Propulsion Engine Dur...

  2. FAST FLUX TEST FACILITY DRIVER FUEL MEETING

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1966-06-01

    The Pacific Northwest Laboratory has convened this meeting to enlist the best talents of our laboratories and industry in soliciting factual, technical information pertinent to the Pacific Northwest's Laboratory's evaluation of the potential fuel systems for the Fast Flux Test Facility. The particular factors emphasized for these fuel systems are those associated with safety, ability to meet testing objectives, and economics. The proceedings includes twenty-three presentations, along with a transcript of the discussion following each, as well as a summary discussion.

  3. Fuel Cell Development and Test Laboratory | Energy Systems Integration

    Science.gov (United States)

    Facility | NREL Fuel Cell Development and Test Laboratory Fuel Cell Development and Test Laboratory The Energy System Integration Facility's Fuel Cell Development and Test Laboratory supports fuel cell research and development projects through in-situ fuel cell testing. Photo of a researcher running

  4. High-temperature distillation and consolidation of U–Zr cathode product from molten salt electrorefining of simulated metallic fuel

    International Nuclear Information System (INIS)

    Iizuka, Masatoshi; Akagi, Masaaki; Koyama, Tadafumi

    2014-01-01

    High-temperature distillation experiments were performed using U–Zr cathode products of various compositions to obtain knowledge on suitable operation conditions and equipment design such as the container material. The LiCl–KCl–UCl 3 electrolyte adhering to the U–Zr cathode products was almost completely vaporized at 1273–1573 K, under pressure of 10–300 Pa. Massive ingots were obtained from the remaining cathode products by heating them at 1573–1673 K. Three different phases were identified in a distillation product of a higher Zr content. A U-rich bulk (3.9 wt% Zr) and a deposit of a relatively low Zr content (17.2 wt% Zr) were considered to be formed during the cooling process of the distillation product. Another Zr-rich deposit (64.7 wt% Zr), which might cause the inhomogeneity of product ingots, was expected to result from Zr-rich spots that originally existed in the cathode product. The Cl content in the cathode product was decreased by distillation to less than 1/200 of that after electrorefining, while it was markedly larger at a higher Zr concentration. To limit the amount of Zr-rich deposit and the Cl content, the amount of Zr in the distillation product should be controlled to a sufficiently low level by optimization of the operating procedures and conditions in the electrorefining and distillation steps. The zirconia coating material developed in this study showed superior performance in inhibiting reaction between the melted U–Zr alloy melt and the graphite crucible and also in the easy release of the U–Zr ingot from the crucible

  5. Reusable fuel test assembly for the FFTF

    International Nuclear Information System (INIS)

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

    1992-01-01

    A fuel test assembly that provides re-irradiation capability after interim discharge and reconstitution of the test pin bundle has been developed for use in the Fast Flux Test Facility (FFTF). This test vehicle permits irradiation test data to be obtained at multiple exposures on a few select test pins without the substantial expense of fabricating individual test assemblies as would otherwise be required. A variety of test pin types can be loaded in the reusable test assembly. A reusable test vehicle for irradiation testing in the FFTF has long been desired, but a number of obstacles previously prevented the implementation of such an experimental rig. The MFF-8A test assembly employs a 169-pin bundle using HT-9 alloy for duct and cladding material. The standard driver pins in the fuel bundle are sodium-bonded metal fuel (U-10 wt% Zr). Thirty-seven positions in the bundle are replaceable pin positions. Standard MFF-8A driver pins can be loaded in any test pin location to fill the bundle if necessary. Application of the MFF-8A reusable test assembly in the FFTF constitutes a considerable cost-saving measure with regard to irradiation testing. Only a few well-characterized test pins need be fabricated to conduct a test program rather than constructing entire test assemblies

  6. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  7. A review of the FT distillate pathway in GHGenius

    International Nuclear Information System (INIS)

    2006-01-01

    Fischer-Tropsch (FT) distillates products are now used throughout the world as a commercial fuel. However, lifecycle greenhouse gas (GHG) emissions from FT distillates fuels are higher than diesel fuel produced from crude oil. This paper provided details of a lifecycle analysis of FT distillates pathways created for GHGenius, a model used to analyze emissions from a variety of combustion sources. The study examined values reported in tests conducted by major oil and gas operators and described the conversion technologies typically used at FT distillates production facilities. Summaries of reports on FT distillates emissions were also provided. Three primary factors were identified that contributed to different results reported for FT distillates emissions: (1) the efficiency of the conversion process; (2) the allocation procedure used in the conventional oil refinery for the emissions of individual products; and (3) the emissions associated with natural gas production. The GHGenius model was used to quantify the impact of the 3 main factors. An alternative system expansion methodology was used to compare crude oil diesel pathways with a high efficiency, low gas leak scenario in order to achieve high values reported by some oil and gas operators. 5 refs., 5 tabs., 5 figs

  8. Test plan for spent fuel cladding containment credit tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1983-11-01

    Lawrence Livermore National Laboratory has chosen Westinghouse Hanford Company as a subcontractor to assist them in determining the requirements for successful disposal of spent fuel rods in the proposed Nevada Test Site repository. An initial scoping test, with the objective of determining whether or not the cladding of a breached fuel rod can be given any credit as an effective barrier to radionuclide release, is described in this test plan. 8 references, 2 figures, 4 tables

  9. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2014-01-01

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

  10. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  11. Fluid flow test for KMRR fuel assemblies

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu; Chung, Chang Hwan; Chun, See Young; Song, Chul Hha; Jun, Hyung Gil; Chung, Heung Joon; Won, Soon Yeun; Cho, Young Rho; Kim, Bok Deuk

    1991-01-01

    Hydraulic and velocity measurment tests were carried out for the KMRR fuel assembly. Two types of the KMRR fuel assembly are consist of longitudinally finned rods. Experimental data of the pressure drops and friction factors for the KMRR fuel assemlby were produced. The measurement technique for the turbulent flow structure in subchannels using the LDV was obtained. The measurement of the experimental constant of the thermal hydraulic analysis code was investigated. The results in this study are used as the basic data for the development of an analysis code. The measurement technique acquired in this study can be applied to the KMRR thermal hydraulic commissioning test and development of the domestic KMRR fuel fabrication. (Author)

  12. Reliability testing of failed fuel location system

    International Nuclear Information System (INIS)

    Vieru, G.

    1996-01-01

    This paper presents the experimental reliability tests performed in order to prove the reliability parameters for Failed Fuel Location System (FFLS), equipment used to detect in which channel of a particular heat transport loop a fuel failure is located, and to find in which channel what particular bundle pair is failed. To do so, D20 samples from each reactor channel are sequentially monitored to detect a comparatively high level of delayed neutron activity. 15 refs, 8 figs, 2 tabs

  13. Using comprehensive two-dimensional gas chromatography for the analysis of oxygenates in middle distillates I. Determination of the nature of biodiesels blend in diesel fuel.

    Science.gov (United States)

    Adam, Frédérick; Bertoncini, Fabrice; Coupard, Vincent; Charon, Nadège; Thiébaut, Didier; Espinat, Didier; Hennion, Marie-Claire

    2008-04-04

    In the current energetic context (increasing consumption of vehicle fuels, greenhouse gas emission etc.) government policies lead to mandatory introduction in fossil fuels of fuels resulting from renewable sources of energy such as biomass. Blending of fatty acid alkyl esters from vegetable oils (also known as biodiesel) with conventional diesel fuel is one of the solutions technologically available; B5 blends (up to 5%w/w esters in fossil fuel) are marketed over Europe. Therefore, for quality control as well as for forensic reasons, it is of major importance to monitor the biodiesel origin (i.e. the fatty acid ester distribution) and its content when it is blend with petroleum diesel. This paper reports a comprehensive two-dimensional gas chromatography (GC x GC) method that was developed for the individual quantitation of fatty acid esters in middle distillates matrices. Several first and the second dimension columns have been investigated and their performances to achieve (i) a group type separation of hydrocarbons and (ii) individual identification and quantitation of fatty acid ester blend with diesel are reported and discussed. Finally, comparison of quantitative GC x GC results with reference methods demonstrates the benefits of GC x GC approach which enables fast and reliable individual quantitation of fatty acid esters in one single run. Results show that under developed chromatographic conditions, quantitative group type analysis of hydrocarbons is also possible, meaning that simultaneous quantification of hydrocarbons and fatty acid esters can be achieved in one single run.

  14. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    Lundberg, L.B.; Croson, M.L.

    1994-01-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  15. 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...

  16. Shale distillation

    Energy Technology Data Exchange (ETDEWEB)

    Blanding, F H

    1948-08-03

    A continuous method of distilling shale to produce valuable hydrocarbon oils is described, which comprises providing a fluidized mass of the shale in a distillation zone, withdrawing hydrocarbon vapors containing shale fines from the zone, mixing sufficient fresh cold shale with the hydrocarbon vapors to quench the same and to cause condensation of the higher boiling constituents thereof, charging the mixture of vapors, condensate, and cold shale to a separation zone where the shale is maintained in a fluidized condition by the upward movement of the hydrocarbon vapors, withdrawing condensate from the separation zone and recycling a portion of the condensate to the top of the separation zone where it flows countercurrent to the vapors passing therethrough and causes shale fines to be removed from the vapors by the scrubbing action of the condensate, recovering hydrocarbon vapors and product vapors from the separation zone, withdrawing preheated shale from the separation zone and charging it to a shale distillation zone.

  17. Tests on CANDU fuel elements sheath samples

    International Nuclear Information System (INIS)

    Ionescu, S.; Uta, O.; Mincu, M.; Prisecaru, I.

    2016-01-01

    This work is a study of the behavior of CANDU fuel elements after irradiation. The tests are made on ring samples taken from fuel cladding in INR Pitesti. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory. By metallographic and ceramographic examination we determinate that the hydride precipitates are orientated parallel to the cladding surface. A content of hydrogen of about 120 ppm was estimated. After the preliminary tests, ring samples were cut from the fuel rod, and were subject of tensile test on an INSTRON 5569 model machine in order to evaluate the changes of their mechanical properties as consequence of irradiation. Scanning electron microscopy was performed on a microscope model TESCAN MIRA II LMU CS with Schottky FE emitter and variable pressure. The analysis shows that the central zone has deeper dimples, whereas on the outer zone, the dimples are tilted and smaller. (authors)

  18. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    Pace, Brett W.; Marinak, Edward A.

    1999-01-01

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U 3 Si 2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  19. New phenomena observed during fuel assemblies testing

    International Nuclear Information System (INIS)

    Tzotcheva, V.

    2001-01-01

    The paper presents a new attempt to explain inexplicable increase of specific activity for some of the fuel assemblies during the fuel tightness testing procedures on Kozloduy NPP. A brief description of established procedure for fuel tightness control is presented in the paper. Special emphasis is given on a hypothesis that explains the fact of existence of deviation in Iodine activity more than usual, which have no reasonable interpretation. The reasons for uniform high Iodine activity for reloaded assemblies, that have kept in the open measuring can for a long time (1-3 hours), is found to be the process of Iodine dissolving in the water and the accelerated process of natural degassing. A proposal to use the 134 Cs and 137 Cs as stand-alone criteria for more precise results is made in respect to increase the reliability of fuel reloading and storage procedures

  20. Corrosion Tests of LWR Fuels - Nuclide Release

    International Nuclear Information System (INIS)

    P.A. Finn; Y. Tsai; J.C. Cunnane

    2001-01-01

    Two BWR fuels [64 and 71 (MWd)/kgU], one of which contained 2% Gd, and two PWR fuels [30 and 45 (MWd)/kgU], are tested by dripping groundwater on the fuels under oxidizing and hydrologically unsaturated conditions for times ranging from 2.4 to 8.2 yr at 90 C. The 99 Tc, 129 I, 137 Cs, 97 Mo, and 90 Sr releases are presented to show the effects of long reaction times and of gadolinium on nuclide release. This investigation showed that the five nuclides at long reaction times have similar fractional release rates and that the presence of 2% Gd reduced the 99 Tc cumulative release fraction by about an order of magnitude over that of a fuel with a similar burnup

  1. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Bell, T

    1863-09-14

    Shales and other bituminous minerals are distilled in horizontal retorts arranged side by side and with furnaces beneath their front ends. The furnace gases pass, preferably through a brickwork grating, into spaces between the retorts and beneath a horizontal partition towards the back ends. They return above the partition to the front of the retorts, and finally enter a horizontal flue leading to a chimney. The front end of each retort is fitted with a hopper for charging and with a door for discharging. The products of distillation pass through perforated partitions inside the retorts and are conveyed away by pipes at the back.

  2. Severe fuel-damage scoping test performance

    International Nuclear Information System (INIS)

    Gruen, G.E.; Buescher, B.J.

    1983-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle

  3. Alpha Fuels Environmental Test Facility impact gun

    International Nuclear Information System (INIS)

    Anderson, C.G.

    1978-01-01

    The Alpha Fuels Environmental Test Facility (AFETF) impact gun is a unique tool for impact testing 238 PuO 2 -fueled heat sources of up to 178-mm dia at velocities to 300 m/s. An environmentally-sealed vacuum chamber at the muzzle of the gun allows preheating of the projectile to 1,000 0 C. Immediately prior to impact, the heat source projectile is completely sealed in a vacuum-tight catching container to prevent escape of its radioactive contents should rupture occur. The impact velocity delivered by this gas-powered gun can be regulated to within +-2%

  4. RECH-1 test fuel irradiation status report

    International Nuclear Information System (INIS)

    Marin, J.; Lisboa, J.; Olivares, L.; Chavez, J.

    2005-01-01

    Since May 2003, one RECH-1 fuel element has been submitted to irradiation at the HFR-Petten, Holland. By November 2004 the irradiation has achieved its pursued goal of 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. This report presents the objectives and the current results of this fuel qualification under irradiation. Besides, a brief description of CHI/4/021, IAEA's Technical Cooperation Project that has supported this irradiation test, is also presented here. (author)

  5. Fuels and materials testing capabilities in Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

    1989-01-01

    The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

  6. Testing of FFTF fuel handling equipment

    International Nuclear Information System (INIS)

    Coleman, D.W.; Grazzini, E.D.; Hill, L.F.

    1977-07-01

    The Fast Flux Test Facility has several manual/computer controlled fuel handling machines which are exposed to severe environments during plant operation but still must operate reliably when called upon for reactor refueling. The test programs for two such machines--the Closed Loop Ex-Vessel Machine and the In-Vessel Handling Machine--are described. The discussion centers on those areas where design corrections or equipment repairs substantiated the benefits of a test program prior to plant operation

  7. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Irvine, R

    1884-04-04

    In obtaining paraffin by distillation of shales, etc., containing sulfur, the steam used for heating is charged with ammonia or ammonium carbonate in suspension. This prevents the sulfur from decomposing the paraffin. The ammonia, etc., may also be used alone or in solution in water.

  8. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C A

    1906-05-22

    The invention relates to an apparatus in which the destructive distillation or coking of coal, peat, shale, etc., is carried out by means of a current of hot gases at a temperature of 700--800/sup 0/F., as described in Specification No. 11,925, A.D. 1906.

  9. Testing system for a fuel cells stack

    International Nuclear Information System (INIS)

    Culcer, Mihai; Iliescu, Mariana; Stefanescu, Ioan; Raceanu, Mircea; Enache, Adrian; Lazar, Roxana Elena

    2006-01-01

    Hydrogen and electricity together represent one of the most promising ways to realize sustainable energy, whilst fuel cells provide the most efficient conversion devices for converting hydrogen and possibly other fuels into electricity. Thus, the development of fuel cell technology is currently being actively pursued worldwide. Due to its simple operation and other fair characteristics, the Proton Exchange Membrane Fuel Cell (PEMFC) is especially suitable as a replacement for the internal combustion engine. The PEMFC is also being developed for decentralized electricity and heat generation in buildings and mobile applications. Starting with 2001 the Institute of Research - Development for Cryogenics and Isotopic Technologies - ICIT - Rm. Valcea developed research activities supported by the Romanian Ministry of Education and Research within the National Research Program in order to bridge the gap to European competencies in the area of hydrogen and fuel cells. The paper deals with the testing system designed and developed in ICIT Rm. Valcea as a flexible and versatile tool allowing a large scale of parameter settings and measurements on a single cell or on a fuel cells stack onto a wind range of output power values. (authors)

  10. Mechanical test for fuel assembly spacer grid

    International Nuclear Information System (INIS)

    Kang, Heung Seok; Jeong, Yeon Ho; Song, Kee Nam; Kim, Hyung Kyu; Yoon, Kyung Ho; Bang, Je Keun.

    1997-06-01

    In order to propose some tests for a new spacer grid, the grid mechanical tests performed by ABB-CE, KWU and Westinghouse have been investigated. It is known that a static compression test, a dynamic impact test, and a grid spring characteristic test were commonly carried out by the vendors when a prototype spacer grid was developed. The static compression test is to measure the stresses on the strips as well as to obtain the grid stiffness. The dynamic impact test is to get some basic data for accident analysis such as impact stiffness, impact strength, and coefficient of restitution. Since each fuel vendor has his theory on an accident analysis, every vendor employs his particular method for the dynamic impact test. The dynamic impact test can be divided into two in accordance with the number of impact face, and the duration of impact pulse. One is an one-sided impact test and the other is an through-gird impact test. The duration of the impact pulse for the former is considerably shorter than the latter. Therefore, the grid can endure much higher load under the one-sided impact condition than under the through-grid impact condition. The grid spring characteristic test is to obtain a force versus deflection curve. This curve is very important in designing the spacer grid to provide fuel rods with a sound supports in core. (author). 18 tabs., 26 figs

  11. Parametric Sensitivity Tests- European PEM Fuel Cell Stack Test Procedures

    DEFF Research Database (Denmark)

    Araya, Samuel Simon; Andreasen, Søren Juhl; Kær, Søren Knudsen

    2014-01-01

    performed based on test procedures proposed by a European project, Stack-Test. The sensitivity of a Nafion-based low temperature PEMFC stack’s performance to parametric changes was the main objective of the tests. Four crucial parameters for fuel cell operation were chosen; relative humidity, temperature......As fuel cells are increasingly commercialized for various applications, harmonized and industry-relevant test procedures are necessary to benchmark tests and to ensure comparability of stack performance results from different parties. This paper reports the results of parametric sensitivity tests......, pressure, and stoichiometry at varying current density. Furthermore, procedures for polarization curve recording were also tested both in ascending and descending current directions....

  12. Fuel tank integrity research : fuel tank analyses and test plans

    Science.gov (United States)

    2013-04-15

    The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. Fuel tank research is being performed to : determine strategies for increasing the fuel tank impact : resistance to ...

  13. Blunt impact tests of retired passenger locomotive fuel tanks

    Science.gov (United States)

    2017-08-01

    The Transportation Technology Center, Inc. conducted impact tests on three locomotive fuel tanks as part of the Federal Railroad Administrations locomotive fuel tank crashworthiness improvement program. Three fuel tanks, two from EMD F40PH locomot...

  14. 14 CFR 25.952 - Fuel system analysis and test.

    Science.gov (United States)

    2010-01-01

    ... using the airplane fuel system or a test article that reproduces the operating characteristics of the... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.952 Fuel system...

  15. Geomechanics of the Spent Fuel Test: Climax

    International Nuclear Information System (INIS)

    Wilder, D.G.; Yow, J.L. Jr.

    1987-07-01

    Three years of geomechanical measurements were made at the Spent Fuel Test-Climax (SFT-C) 1400 feet underground in fractured granitic rock. Heating of the rock mass resulted from emplacement of spent fuel as well as the heating by electrical heaters. Cooldown of the rock occurred after the spent fuel was removed and the heaters were turned off. The measurements program examines both gross and localized responses of the rock mass to thermal loading, to evaluate the thermomechanical response of sheared and fractured rock with that of relatively unfractured rock, to compare the magnitudes of displacements during mining with those induced by extensive heating of the rock mass, and to check assumptions regarding symmetry and damaged zones made in numerical modeling of the SFT-C. 28 refs., 113 figs., 10 tabs

  16. Corrosion testing of uranium silicide fuel specimens

    International Nuclear Information System (INIS)

    Bourns, W.T.

    1968-09-01

    U 3 Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300 o C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U 3 5i specimen which corrodes at less than 2 mg/cm 2 h in 300 o C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U 3 Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300 o C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  17. Iowa Central Quality Fuel Testing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Heach, Don; Bidieman, Julaine

    2013-09-30

    The objective of this project is to finalize the creation of an independent quality fuel testing laboratory on the campus of Iowa Central Community College in Fort Dodge, Iowa that shall provide the exploding biofuels industry a timely and cost-effective centrally located laboratory to complete all state and federal fuel and related tests that are required. The recipient shall work with various state regulatory agencies, biofuel companies and state and national industry associations to ensure that training and testing needs of their members and American consumers are met. The recipient shall work with the Iowa Department of Ag and Land Stewardship on the development of an Iowa Biofuel Quality Standard along with the Development of a standard that can be used throughout industry.

  18. Corrosion testing of uranium silicide fuel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Bourns, W T

    1968-09-15

    U{sub 3}Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300{sup o}C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U{sub 3}5i specimen which corrodes at less than 2 mg/cm{sup 2} h in 300{sup o}C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U{sub 3}Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300{sup o}C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  19. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Neilson, A; Renfrewshire, I; Black, W

    1889-06-14

    The invention relates to a method of, and apparatus for, distilling shale, coal, or other oil or tar-yielding minerals, to obtain gases, liquids, or other products. The distillation is effected in vertical retorts by the combustion of the partially spent material in the lower part of the retorts, to which steam and air are admitted. The retorts are built of firebrick, and provided with iron casings. They are fed through hoppers and discharged through the openings. The discharging is facilitated by a cone, or its equivalent, in the base of each retort. Steam and air are admitted through the pipes. The interior may be viewed through holes. The products are taken off from the space around the hopper.

  20. Distilling shale

    Energy Technology Data Exchange (ETDEWEB)

    Armour, J; Armour, H

    1889-05-07

    The invention relates to retorts and accessory apparatus for distilling shale or other oil-yielding minerals. A series of long vertical retorts, composed of fire-brick or similar refractory material, are arranged in two rows in a bench, being divided into groups of four by transverse vertical partitions. The retorts are surmounted by metal casings or hoppers into which the fresh mineral is charged, and from which the distillate passes off through lateral pipes. Any uncondensed gases from the retorts may be passed into the flues surrounding them by the pipe and burned. The products of combustion from a furnace pass through a series of horizontal flues, being compelled to pass completely round each retort before entering the flue above. The products from two or more sets pass from the upper flues into flues running along the top of the bench, and return through a central flue to the chimney.

  1. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Urquhart, D

    1882-08-19

    Manufacture of ammonia and purification of shale oils are described. In the distillation of shales, increase of ammonia is obtained and the oils are less contaminated by sulfur, by mixing a small proportion, about two to three percent, of lime or slaked line with the shale. The same process is used with other carbonaceous substances containing nitrogen, such as peat, coal, or the like; but a smaller proportion of lime is used than in the former case, and the lime is slaked with caustic soda solution. When slack or waste coal or other carbonaceous substances are distilled by heated air or gases arising from imperfect combustion, as in furnaces on the gas producer principle, slaked lime is added to the slack or other material.

  2. Distilling oils

    Energy Technology Data Exchange (ETDEWEB)

    Leffer, L G

    1912-01-29

    In a process for converting heavy hydrocarbons, such as petroleum or shale oil, into light hydrocarbons by distilling under the pressure of an inert gas, the operation is conducted at a temperature not exceeding 410/sup 0/C and under an accurately regulated pressure, the gas being circulated through the still and the condenser by means of a pump. The oil in the still may be agitated by stirring vanes or by blowing the gas through it. Hydrogen, nitrogen, carbon dioxide, methane, or gases generated in the distillation may be used as pressure media; the gas is heated before its admission to the still. A pressure of from 11 to 12 atmospheres is used in treating gas oil. Specification 10,277/89 is referred to.

  3. Distilling shale

    Energy Technology Data Exchange (ETDEWEB)

    Kern, L

    1922-07-21

    In the distillation of shale and similar materials the shale is ground and briquetted and the briquettes are placed in a retort so that air passages are left between them, after which they are uniformly and slowly heated to at least 700/sup 0/C, the air passages facilitating the escape of the oil vapors, and the slow heating preventing fusion of the flux forming constituents. After the bitumen has been driven off, air is passed into the retort and heating continued to about 1050/sup 0/C, the result being a porous product suitable for insulating purposes or as a substitute for kieselguhr. The ground shale may be mixed prior to distillation with peat, sawdust, or the like, and with substances which yield acids, such as chlorides, more particularly magnesium chloride, the acids acting on the bitumen.

  4. Distilling coal

    Energy Technology Data Exchange (ETDEWEB)

    Blythe, F C

    1914-09-14

    In the destructive distillation of bituminous coal, heavy hydrocarbon oil, such as petroleum, kerosine, shale oil, and heavy tar oil, obtained in some cases during the process, is added to the coal, which is then distilled under pressure and at a comparatively low temperature regulated so as to produce a large proportion of hydrocarbon oils and a small proportion of permanent gas. In one method, about 5 to 10 parts of hydrocarbon oil are mixed with 100 parts of crushed or ground coal, and the mixture is heated in a closed vessel, provided in some cases with an agitator, under a pressure of about 60 lb/in/sup 2/, and the temperature may be gradually raised to 350/sup 0/C and then to about 500/sup 0/C. The heating may be by means of superheated steam with or without external heat.

  5. Toxicity of middle distillates from dermal exposure.

    Science.gov (United States)

    Koschier, F J

    1999-02-01

    This report focuses on recent studies that investigated the effects of kerosine dermal exposure on neurotoxicity and reproductive/developmental toxicity. Background toxicity information will also be reviewed for kerosine range mid distillates. The kerosine range mid distillates have a carbon range of C9-C16 and have a boiling range of 302-554 degrees F (150-290 degrees C). This category includes kerosine, aviation fuels (e.g., Jet A, JP-5 and JP-8), no. 1 fuel oil and diesel fuel oil. In general, the kerosine range mid distillates demonstrate relatively low acute toxicity by any route of exposure. High inhalation exposures can induce central nervous system depression characterized by ataxia, hypoactivity and prostration. Kerosines are known to cause skin irritation and inflammation under conditions of acute and repeated exposure in animals and humans, but are only slightly irritating to the eye and are not skin sensitizers. In addition, the absorption of kerosine range mid distillates through the skin has been demonstrated to be fairly rapid, but limited to approximately 10-15% of the applied dose after 24 hours. The kerosine range mid distillates are generally inactive in genetic toxicity tests although positive studies have been reported. Positive results, while at times equivocal, have been reported for straight run kerosine and jet fuel A in the mouse lymphoma assay with metabolic activation, and hydrodesulfurized kerosine (mouse) and jet fuel A (rat) in the bone marrow cytogenetic assay. Effects on the nervous and reproductive systems have been reported in humans and experimental animals under conditions where inhalation and dermal exposure to specific kerosine type fuels are sometimes difficult to separate. Recent laboratory studies have addressed this point and examined the effects of dermal exposure. In these studies, rats were exposed to hydrodesulfurized kerosine by skin application to determine the potential of dermal contact to cause reproductive

  6. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, J M

    1884-06-03

    The invention relates to retorts for the destructive distillation of shale, dross, and other carbonaceous or oleaginous materials, and for the distillation and carbonization of sawdust, shavings, tan bark, and the like. The material is fed from a trapped hopper on a series of trays or casings of cast iron or other material, separated by flue spaces and arranged in a tier round a vertical rotating shaft passing through tubular pieces cast on the casings. The shaft is fitted with arms which carry stirring-blades so disposed that the material is shifted from side to side and slowly fed towards the ducts through which it passes to the casing next below, and is finally withdrawn from the apparatus by a pipe, which may be trapped or otherwise. Furnace gases are admitted through openings in the enclosing brickwork having settings to support the casings, the lowermost of which may be fitted below the inlet for furnace gases and their contents cooled by the circulation of cold water round them. The gaseous or volatile products of distillation pass to a condenser by means of openings and the pipe, which may be formed in sections to obtain access to the casings, or doors may be provided for this purpose. The ducts may be arranged alternately at the edge and center of the casings, which may be jacketed, and heated air or steam may be employed instead of furnace gases. Means may also be provided for admitting superheated steam into one or more of the casings.

  7. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Young, W

    1877-03-29

    The method consists in agitating or circulating the distillation products inside the retort by means of jets of gas, steam, or vapor, or by means of reciprocating pistons; condensing certain of the heavy hydrocarbons; sealing or luting the doors of retorts or distilling-vessels; and conducting the distillation for the manufacture of oil so that the charging or discharging doors may be fitted with self-sealing lids. Several arrangements are shown and described; a single horizontal retort is divided into two compartments by a perforated plate which supports the coal, shale, or other bituminous substance, beneath which a piston is reciprocated or a jet of steam, gas, or vapor injected; a vertical retort is fitted with a central tube into which steam, gas, or vapor is injected, or it may be divided into two compartments and the jet injected into one of these; a pair of vertical retorts are connected by a horizontal passage at the top and bottom, and into the upper one steam, gas, or vapor is injected, or the lower one is fitted with a piston.

  8. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Newton, A V

    1856-04-22

    In order to obtain, at the first distillation, from coal, shale, and bituminous substances an oil sufficiently pure for illuminating and other purposes, the material broken into very small pieces and placed on the bottom of the retort, is evenly covered with common sand, about four times greater in weight than the weight of the coal. The coal and sand are then gradually raised to a temperature of 212/sup 0/F. Steam containing carbonaceous impurities first passes to the condenser, and subsequently oil, which rises to the surface of the water in the receiving-vessel. When some bituminous substances are employed, the temperature, after oil ceases to come over, may be gradually raised until the oil produced ceases to be pure. Most kinds of clay and earth, chalk, gypsum, black oxide of manganese, plumbago, or charcoal may be used separately, in combination, or with added chemicals, instead of sand as the medium for filtering the gas or vapor from which the oil is formed. Either the oil obtained by the first distillation or oils obtained by other means may be rectified by distilling with sand.

  9. Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

    1995-01-01

    This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

  10. Health assessment of gasoline and fuel oxygenate vapors: generation and characterization of test materials.

    Science.gov (United States)

    Henley, Michael; Letinski, Daniel J; Carr, John; Caro, Mario L; Daughtrey, Wayne; White, Russell

    2014-11-01

    In compliance with the Clean Air Act regulations for fuel and fuel additive registration, the petroleum industry, additive manufacturers, and oxygenate manufacturers have conducted comparative toxicology testing on evaporative emissions of gasoline alone and gasoline containing fuel oxygenates. To mimic real world exposures, a generation method was developed that produced test material similar in composition to the re-fueling vapor from an automotive fuel tank at near maximum in-use temperatures. Gasoline vapor was generated by a single-step distillation from a 1000-gallon glass-lined kettle wherein approximately 15-23% of the starting material was slowly vaporized, separated, condensed and recovered as test article. This fraction was termed vapor condensate (VC) and was prepared for each of the seven test materials, namely: baseline gasoline alone (BGVC), or gasoline plus an ether (G/MTBE, G/ETBE, G/TAME, or G/DIPE), or gasoline plus an alcohol (G/EtOH or G/TBA). The VC test articles were used for the inhalation toxicology studies described in the accompanying series of papers in this journal. These studies included evaluations of subchronic toxicity, neurotoxicity, immunotoxicity, genotoxicity, reproductive and developmental toxicity. Results of these studies will be used for comparative risk assessments of gasoline and gasoline/oxygenate blends by the US Environmental Protection Agency. Copyright © 2014 Elsevier Inc. All rights reserved.

  11. CANFLEX fuel bundle cross-flow endurance test (test report)

    International Nuclear Information System (INIS)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

    1997-04-01

    As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs

  12. CANFLEX fuel bundle cross-flow endurance test (test report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

    1997-04-01

    As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

  13. 14 CFR 29.965 - Fuel tank tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 29.965 Section 29.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.965 Fuel tank tests. (a) Each fuel tank...

  14. 14 CFR 27.965 - Fuel tank tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 27.965 Section 27.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.965 Fuel tank tests. (a) Each fuel tank...

  15. Test of high temperature fuel element, (1)

    International Nuclear Information System (INIS)

    Akino, Norio; Shiina, Yasuaki; Nekoya, Shin-ichi; Takizuka, Takakazu; Emori, Koichi

    1980-11-01

    Heat transfer experiment to measure the characteristics of a VHTR fuel in the same condition of the reactor core was carried out using HTGL (High Temperature Helium Gas Loop) and its test section. In this report, the details of the test section, related problems of construction and some typical results are described. The newly developed heater with graphite heat transfer surface was used as a simulated fuel element to determine the heat transfer characteristics. Following conclusions were obtained; (1) Reynolds number between turbulent and transitional region is about 2600. (2) Reynolds number between transitional and laminar region is about 4800. (3) The laminarization phenomena have not been observed and are hardly occurred in annular tubes comparing with round tube. (4) Measured Nusselt numbers agree to the established correlations in turbulent and laminar regions. (author)

  16. LOFT fuel rod surface temperature measurement testing

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

    1978-01-01

    Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

  17. PIE Report on the KOMO-3 Irradiation Test Fuels

    International Nuclear Information System (INIS)

    Park, Jong Man; Ryu, H. J.; Yang, J. H.

    2009-04-01

    In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

  18. Post irradiation examination on test fuel pins for PWR

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Ambrozio Filho, F.

    1981-01-01

    Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

  19. Fuel Retrieval Sub-Project (FRS) Stuck Fuel Station Performance Test Data Report

    International Nuclear Information System (INIS)

    THIELGES, J.R.

    2000-01-01

    This document provides the test data report for Stuck Fuel Station Performance Testing in support of the Fuel Retrieval Sub-Project. The stuck fuel station was designed to provide a means of cutting open a canister barrel to release fuel elements, etc

  20. Test Results of PBMR Fuel Spheres

    International Nuclear Information System (INIS)

    Koshcheev, Konstantin; Diakov, Alexander; Beltyukov, Igor; Barybin, Andrey; Chernetsov, Mikhail

    2014-01-01

    Results of pre-irradiation testing of fuel spheres (FS) and coated particles (CP) manufactured by PBMR SOC (Republic of South Africa) are described. The stable high quality level of major characteristics (dimensions, CP coating structure, uranium-235 contamination of the FS matrix graphite and the outer PyC layer of the CP coating) are shown. Results of a methodical irradiation test of two FS in helium and neon medium at temperatures of 800 to 1300 °C with simultaneous determination of release-to-birth ratios for major gaseous fission products (GFP) are described. (author)

  1. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    1915-07-03

    Oils are extracted from coal, lignite, shale, boghead, butumen, asphalt, tar, pitch, etc., by distillation at a low temperature, which may be 300 to 425/sup 0/CC, solvent oils or vapors being circulated during the heating which may be conducted with or without increased or reduced pressure. The solvent oils and the extracted oils are recovered in condensers, etc., last traces being expelled from the material by a current of water vapor. The uncondensed gases may be used for heating, and the solid residue may be used for the production of gas and coke, or may be briquetted.

  2. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Young, W; Brash, P

    1866-05-04

    In the distillation of oil from coal, shale, etc., hydrocarbon vapors which are condensed only with difficulty, and are of small value, are reheated and sent back into the retorts. A jet of steam, or a forcing or exhausting apparatus, may be used for this purpose, and the vapors are passed under false bottoms with which the retorts are preferably provided. In the rectification of the oils, a producer known as still bottoms results which, when redistilled, gives rise to vapors condensable only with difficulty. These vapors may be passed back into the still, or may be mixed and heated, in a separate vessel, with the vapors coming from the still.

  3. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Medlrum, E

    1869-02-06

    The invention relates to the decomposition of the liquids with high boiling points, and the solids with low melting points, left in the purification of paraffin oil obtained from coal or shale. The liquids or melted solids, or their vapors, are passed through a heated iron tube or retort and c., which may be packed with broken stones, spent shale, and c. The temperature is regulated between 700/sup 0/F and a low red heat. The condensed products consist of a mixture of light and heavy oils, which may be separated by distillation. The heavier residues may be again passed through the decomposing apparatus.

  4. Distilling hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    Bataafsche, N V; de Brey, J H.C.

    1918-10-30

    Hydrocarbons containing a very volatile constituent and less volatile constituents, such as casing-head gases, still gases from the distillation of crude petroleum and bituminous shale are separated into their constituents by rectification under pressure; a pressure of 20 atmospheres and limiting temperatures of 150/sup 0/C and 40/sup 0/C are mentioned as suitable. The mixture may be subjected to a preliminary treatment consisting in heating to a temperature below the maximum rectification temperature at a pressure greater than that proposed to be used in the rectification.

  5. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  6. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  7. Advanced distillation curve measurements for corrosive fluids: Application to two crude oils

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Starkey Ott; Beverly L. Smith; Thomas J. Bruno [National Institute of Standards and Technology, Boulder, CO (United States). Physical and Chemical Properties Division

    2008-10-15

    We have recently introduced several important improvements in the measurement of distillation curves for complex fluids. The modifications include a composition-explicit data channel for each distillate fraction (for both qualitative and quantitative analysis) and corrosivity assessment of each distillate fraction. The composition-explicit information is achieved with a new sampling approach that allows precise qualitative as well as quantitative analyses of each fraction, on the fly. We have applied the new method to a variety of fluids, including simple n-alkanes, rocket propellant, gasoline, jet fuels, and a hydrocarbon fluid made corrosive with dissolved hydrogen sulfide. In the current contribution, we present the application of the advanced distillation curve method to two samples of crude oil. A primary motivation behind the work is to precisely measure the distillation curves of these oils using our advanced distillation apparatus; these low uncertainty measurements of true thermodynamic state points can be used for equation of state development and differentiation of crude oil samples. Then, the information content of each distillation was extended much further by use of the composition-explicit data channel: gas chromatography-mass spectrometry (GC-MS), infrared spectrophotometry (IR), gas chromatography with sulfur chemiluminescence detection (GC-SCD), and the copper strip corrosion test (CSCT) were used for each distillate volume fraction sampled. Consequently, for each volume fraction of crude oil distillate sampled, we can address the composition, quantitate the total sulfur content, and measure the corrosivity. 39 refs., 6 figs., 2 tabs.

  8. Legacy Vehicle Fuel System Testing with Intermediate Ethanol Blends

    Energy Technology Data Exchange (ETDEWEB)

    Davis, G. W.; Hoff, C. J.; Borton, Z.; Ratcliff, M. A.

    2012-03-01

    The effects of E10 and E17 on legacy fuel system components from three common mid-1990s vintage vehicle models (Ford, GM, and Toyota) were studied. The fuel systems comprised a fuel sending unit with pump, a fuel rail and integrated pressure regulator, and the fuel injectors. The fuel system components were characterized and then installed and tested in sample aging test rigs to simulate the exposure and operation of the fuel system components in an operating vehicle. The fuel injectors were cycled with varying pulse widths during pump operation. Operational performance, such as fuel flow and pressure, was monitored during the aging tests. Both of the Toyota fuel pumps demonstrated some degradation in performance during testing. Six injectors were tested in each aging rig. The Ford and GM injectors showed little change over the aging tests. Overall, based on the results of both the fuel pump testing and the fuel injector testing, no major failures were observed that could be attributed to E17 exposure. The unknown fuel component histories add a large uncertainty to the aging tests. Acquiring fuel system components from operational legacy vehicles would reduce the uncertainty.

  9. Process of distilling heavy hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    1929-12-03

    This invention has for its object the distillation of heavy liquid hydrocarbons for the purpose of obtaining lighter hydrocarbons stable and immediately salable for fuels in combustion motors. The process is distinguished by the fact that the heavy hydrocarbon is distilled by means of heating to a temperature in keeping with the nature of the material to be treated up to 350/sup 0/C under pressure or without pressure the distillation being carried out on catalysts containing successively nickel, copper, and iron (3 parts of nickel, 1 part of copper, and 1 part of iron), the vapors produced by this distillation being exposed in turn to the action of catalysts of the same nature and in the same proportion.

  10. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, H; Laing, B

    1927-12-23

    Oil bearing solids such as coal, tar sands, oil shales, or the like, are distilled in a current of hot gas and are first preheated to a temperature above that at which the heaviest oil fractions in the vapors are liable to condense, for example 100 to 250/sup 0/C, according to the volume of gas passing through the retort, temperature being in inverse proportion to the quantity of oxygen containing constituents in the charge. When the distillation takes place in a controlled volume of hot inert gas of 45,000 cubic feet per ton and the volume of oil recovered is about 20 gallons per ton, the material is preheated to 200 to 250/sup 0/C, when the volume of gas used is 100,000 cubic feet the preheating temperature is 150/sup 0/C. The temperatures of the retort dust extractor etc. do not fall below 100 to 150/sup 0/C until actual condensation of the oil vapor is desired. Specification 287,381 is referred to, and Specification 287,037 also is referred to in the Provisional Specification.

  11. Shale distillation

    Energy Technology Data Exchange (ETDEWEB)

    Jacomini, V V

    1938-06-07

    To produce valuable oils from shale by continuous distillation it is preheated by a heated fluid and introduced into a distilling retort from which the oil vapours and spent material are separately removed and the vapours condensed to recover the oil. The shale is preheated to 400 to 500/sup 0/F in the hopper by combustion gases from a flue and is fed in measured quantities to a surge drum, a loading chamber and surge drum, the latter two being connected to a steam pipe which equalises the pressure thereon. The material passes by two screw conveyors to a retort with deflector bars to scatter the material so that lean hot cycling gas flowing through a pipe is spread out as it makes its way upwardly through the shale and heats the oil so that it is driven off as vapour, collected in the lean gas and carried off through an outlet pipe. A measuring valve is provided at the bottom of a retort and cutter knives cut the spent shale and distribute cooling water thereto. The gases travel through heat exchangers and a condenser to an accumulator where the cycling gas is separated from the vapours, passed to compression, and preheated in a gas exchanger and spiral coils before it is returned to the retort. The oil passes to a storage tank by way of a unit tank in which oil vapours are recovered. Water is collected by a pipe in the tank bottom and returned by shaft to a retort.

  12. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Young, W

    1881-04-12

    Destructive distillation of shale for the manufacture of mineral oil and ammonia is described. The retorts are arranged in benches, each retort being placed over its own combustion chamber into which the spent shale is discharged and consumed in heating the next charge as described in Specification No. 1578, A. D. 1880. Two forms of retorts are shown, each consisting of two retorts placed above and communicating with one another, the upper being employed to distill the oil at a low red heat, and the lower to eliminate the nitrogen in the form of ammonia at a much higher temperature. The retorts are divided by a sliding damper and have an outlet for the passage of the products placed at the junction. The retorts have an outlet at the top for the escape of the products. Each retort has an opening closed by a cover for charging and a door for discharging. The products of combustion from the combustion chambers pass through ports to a chamber surrounding the lower retorts and thence through ports in the division wall controlled by dampers into the chamber surrounding the upper retorts, whence they pass through flues to the chimney. Around the bottom of each retort are openings communicating with a chamber to which steam is admitted through a valve from a pipe preferably placed in a coil in the flue.

  13. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Byrom, W A; Bennett, J A.B.

    1884-02-22

    Effecting the separation of the various products of the distillation of coal, shale, etc., by taking advantage of the graduated temperature of successive portions of the length of mechanical retorts is described. The substances entering from the hopper are gradually heated in their progress and give off a series of products in their order of volatilization, which pass from openings in the retort through a series of ascension pipes into collecting-vessels. The vessels are designed to contain different portions of the distillate and are sealed against the escape of uncondensed vapor or gas by the condensed liquid. Each of the ascension pipes communicates above its vessel with a common pipe to convey away permanent gases. The flues for heating the retort may be so arranged as to give the greatest heat at the end farthest from the point of entrance, or the stages of heat may be self-regulated by the time necessary for the material to acquire heat as it travels. If necessary the pipes may be fitted with refrigerating-appliances.

  14. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Mitting, E K

    1882-08-09

    The broken-up shale is placed in cast-iron retorts, heated externally, having exit tubes placed at a low level. Each retort is provided with a steam-pipe with a regulating-cock outside, the pipe being carried around the walls of the retort in a spiral or zig-zag way to ensure superheating of the steam, perforations being made in the pipe to allow exit for the steam into the retort. The steam, which may if desired be superheated before entrance, is passed into the retort when the latter has attained a temperature of from 210 to 250/sup 0/C and the passage is continued while the temperature rises, as long as distillation goes on. The exit pipe to the retort leads to a condenser of much condensing-surface, provided with a drag obtained by an exhausting steam jet or otherwise. The distilled products consist of tar, oils, wax, ammoniacal water (stated to be in greater proportion through the use of the process), and lighting and heating gas. The latter gas goes through a scrubber to a gasholder. The carbonaceous residue in the retort is discharged when cooled below a red heat, into sheet-iron cylinders, with tightly fitting lids, to avoid as far as possible contact with the atmosphere.

  15. Spent fuel cladding containment credit test

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1983-01-01

    As an initial step in addressing the effectiveness of breached cladding as a barrier to radionuclide release from the repository during the post-containment period, preliminary scoping tests have been initiated which compare radionuclide releases from spent fuel specimens with artificially induced cladding defects of various severities. The artificially induced defects are all more severe than the typical in-reactor type breaches which are expected to be the principal type of breach entering the repository for terminal storage. These preliminary scoping tests being conducted by Westinghouse Hanford Company for the Lawrence Livermore National Laboratory Waste Package Development Program in support of the Tuff repository project at the Nevada Test Site are described. Also included in this presentation are selected initial results from these tests. 22 figures

  16. Design data, liquid distributors and condenser for a distillation column to enrich tritium in metallic lithium

    International Nuclear Information System (INIS)

    Barnert, E.

    1984-01-01

    Tritium, one fuel component of the fusion reactor is bred from the reactors blanket material lithium. Before extracting the tritium from, for instance, metallic lithium by permeation it has to be enriched in the lithium by high temperature distillation. In this report design data for a column for high temperature distillation are given. About the testing of two distributors for small liquid quantities and of a condenser is reported. (orig.) [de

  17. Method and arrangement of distillation of shales

    Energy Technology Data Exchange (ETDEWEB)

    Bergh, S V

    1920-03-29

    A method is given of distilling shale and other bituminous materials utilizing the heat from the combustion of the residue, possibly with additional heat from other fuels. It is characterized by the shale, which is arranged in layers, being first submitted to a process of distillation utilizing the heat mentioned, and at the same time recovering the products of distillation, and second the shale being burned without disturbing the layers to any appreciable extent. The patent has 16 more claims.

  18. F2 phenomenological test on fuel motion (Interim report)

    International Nuclear Information System (INIS)

    Palm, R.G.; Fink, C.L.; Stewart, R.R.; Gehl, S.M.; Rothman, A.B.

    1976-09-01

    TREAT F-series tests are being conducted to provide data on fuel motion at accident power levels from one to about ten times design for use in development of fuel motion models. Test F2 was conducted to evaluate motion of high power fuel in a hypothetical LMFBR unprotected TUC (transient undercooling) accident. Fuel and fuel-boundary conditions following coolant boiling and dryout under TUC conditions are achieved in each F-series test with a single fuel element surrounded by a nuclear heated wall in a dry test capsule. Test F2 was conducted with a low burnup but restructured fuel element to investigate the effect of fuel vapor pressure on fuel motion. Results are presented and discussed

  19. Fuel fragmentation data review and separate effects testing

    International Nuclear Information System (INIS)

    Yueh, Ken. H.; Snis, N.; Mitchell, D.; Munoz-Reja, C.

    2014-01-01

    A simple alternative test has been developed to study the fuel fragmentation process at loss of coolant accident (LOCA) temperatures. The new test heats a short section of fuel, approximately two pellets worth of material, in a tube furnace open to air. An axial slit is cut in the test sample cladding to reduce radial restraint and to simulate ballooned condition. The tube furnace allows the fuel fragmentation process be observed during the experiment. The test was developed as a simple alternative so large number of tests could be conducted quickly and efficiently to identify key variables that influence fuel fragmentation and to zeroing on the fuel fragmentation burn-up threshold. Several tests were conducted, using fuel materials from fuel rods that were used in earlier integral tests to benchmark and validate the test technique. High burn-up fuel materials known to be above the fragmentation threshold was used to evaluate the fragmentation process as a function of temperature. Even with an axial slit and both ends open, no significant fuel detachment/release was detected until above 750°C. Additional tests were conducted with fuel materials at burn-ups closer to the fuel fragmentation burn-up threshold. Results from these tests indicate a minor power history effect on the fuel fragmentation burn-up threshold. An evaluation of available literature and data generated from this work suggest a fuel fragmentation burn-up threshold between 70 and 75 GWd/MTU. (author)

  20. Liquefied Gaseous Fuels Spill Test Facility

    International Nuclear Information System (INIS)

    1993-02-01

    The US Department of Energy's liquefied Gaseous Fuels Spill Test Facility is a research and demonstration facility available on a user-fee basis to private and public sector test and training sponsors concerned with safety aspects of hazardous chemicals. Though initially designed to accommodate large liquefied natural gas releases, the Spill Test Facility (STF) can also accommodate hazardous materials training and safety-related testing of most chemicals in commercial use. The STF is located at DOE's Nevada Test Site near Mercury, Nevada, USA. Utilization of the Spill Test Facility provides a unique opportunity for industry and other users to conduct hazardous materials testing and training. The Spill Test Facility is the only facility of its kind for either large- or small-scale testing of hazardous and toxic fluids including wind tunnel testing under controlled conditions. It is ideally suited for test sponsors to develop verified data on prevention, mitigation, clean-up, and environmental effects of toxic and hazardous gaseous liquids. The facility site also supports structured training for hazardous spills, mitigation, and clean-up. Since 1986, the Spill Test Facility has been utilized for releases to evaluate the patterns of dispersion, mitigation techniques, and combustion characteristics of select materials. Use of the facility can also aid users in developing emergency planning under US P.L 99-499, the Superfund Amendments and Reauthorization Act of 1986 (SARA) and other regulations. The Spill Test Facility Program is managed by the US Department of Energy (DOE), Office of Fossil Energy (FE) with the support and assistance of other divisions of US DOE and the US Government. DOE/FE serves as facilitator and business manager for the Spill Test Facility and site. This brief document is designed to acquaint a potential user of the Spill Test Facility with an outline of the procedures and policies associated with the use of the facility

  1. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  2. Ignition delay times of Gasoline Distillation Cuts measured with Ignition Quality Tester

    KAUST Repository

    Naser, Nimal

    2017-04-21

    Tailoring fuel properties to maximize the efficiency of internal combustion engines is a way towards achieving cleaner combustion systems. In this work, the ignition properties of various gasoline fuel distillation cuts are analyzed to better understand fuel properties of the full boiling range fuel. An advanced distillation column (ADC) provides a more realistic representation of volatility characteristics, which can be modeled using equilibrium thermodynamic methods. The temperature reported is that of the liquid, as opposed to the vapor temperature in conventional ASTM D86 distillation standard. Various FACE (fuels for advanced combustion engines) gasolines were distilled and various cuts were obtained. The separated fractions were then tested in an ignition quality tester (IQT) to see the effect of chemical composition of different fractions on their ignition delay time. Fuels with lower aromatic content showed decreasing ignition delay time with increasing boiling point (i.e., molecular weight). However, fuels with higher aromatic content showed an initial decrease in ignition delay time with increasing boiling point, followed by drastic increase in ignition delay time due to fractions containing aromatics. This study also provides an understanding on contribution of different fractions to the ignition delay time of the fuel, which provides insights into fuel stratification utilized in gasoline compression ignition (GCI) engines to tailor heat release rates.

  3. Test of irradiation of tellurium oxide for obtaining iodine-131 by dry distillation

    International Nuclear Information System (INIS)

    Alanis M, J.

    2003-07-01

    With the purpose of optimizing to the maximum independently the work of the reactor of those mathematical calculations of irradiation that are already optimized, now it corresponds to carry out irradiation tests in the different positions with their respective neutron fluxes that it counts the reactor for samples irradiation. Then, it is necessary to carry out the irradiation of the tellurium dioxide through cycles, with the purpose of observing the activity that it goes accumulating in each cycle and this way to obtain an activity of the Iodine-131 obtained when finishing the last cycle. (Author)

  4. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  5. Preliminary nuclear design for test MOX Fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

    1997-10-01

    As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

  6. Becker's rational addiction theory: An empirical test with price elasticities for distilled spirits in Denmark 1911-31.

    Science.gov (United States)

    Skog, Ole-Jørgen; Melberg, Hans Olav

    2006-10-01

    To test an implication of Becker's rational addiction theory, namely that price changes will lead both to simultaneous consumption changes as well as lagged changes (and potentially also immediate changes if future changes in prices are anticipated). Time-series analysis, first of aggregate sales of distilled spirits and prices, controlled for gross national product (GNP), and secondly of deaths from delirium tremens. Denmark 1911-31. Price changes were very large in the period 1916-18 due to shortages during World War I, and the Danish case can be conceived as a natural experiment. No evidence for lagged price effects in the expected direction was found. On the contrary, the evidence pointed in the opposite direction. The immediate reduction in sales following rising prices are, to some degree, counteracted by an adjustment in the opposite direction the following year. The delirium tremens data confirm this pattern. Becker's theory is not confirmed. Several possible explanations are discussed. If the pattern observed in these data is representative of a more general mechanism, current price elasticity estimates may be too high, by ignoring lagged compensatory effects.

  7. Putting to point the production process of iodine-131 by dry distillation (Preoperational tests)

    International Nuclear Information System (INIS)

    Alanis M, J.

    2002-12-01

    With the purpose of putting to point the process of production of 131 I, one of the objectives of carrying out the realization of operational tests of the production process of iodine-131, it was of verifying the operation of each one of the following components: heating systems, vacuum system, mechanical system and peripheral equipment that are part of the production process of iodine-131, another of the objectives, was settling down the optimal parameters that were applied in each process during the obtaining of iodine-131, it is necessary to point out that this objective is very important, since the components of the equipment are new and its behavior during the process is different to the equipment where its were carried out the experimental studies. (Author)

  8. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Tennent, R B

    1886-12-02

    The invention has reference to an improved construction or mode of building, arranging, and combining the parts of gas-heated retorts for the distillation of shale and other minerals, which by the improved mode and means comprised therein of heating the retorts by the gases, combined with highly heated air and the use of superheated steam in the retorts, and the utilization of the heat of the escaping waste gases for the superheating of the steam, and the raising of steam in boilers for motive power and other purposes. The retorts are erected in transverse pairs, each retort having its surrounding flame flues heating the air for each pair and with steam superheating chambers and pipes between for each pair heated by the escaping gases from the retorts.

  9. Destructive, distillation

    Energy Technology Data Exchange (ETDEWEB)

    Jameson, J

    1882-10-23

    The apparatus employed resembles a reverberatory furnace, having a brickwork chamber with pipes or passages leading from the bottom, through which gases and vapors, arising from destructive distillation or heating of the materials with which the chamber is charged to a certain depth, are drawn by suction produced by a fan or blower. The materials are heated from above by firegates admitted from a separate furnace or fireplace. When shale is thus treated, to obtain burning gas, oil, and ammonia, the suction may be so regulated as to give preponderance to whichever product is desired, the depth of material treated being also concerned in the result. The process is applicable also in the treatment of coal pit refuse, sawdust, peat, and other matters, to obtain volatile products; in burning limestone to obtain carbon dioxide; and in roasting ores. Reference is made to a former Specification for coking coal, No. 1947, A. D. 1882.

  10. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Craig, A

    1863-11-12

    To obtain hydrocarbons, coal, shale, or other bituminous substance is distilled in an annular retort. The outer surface is heated by an arrangement of furnaces and flues, and the inner surface has a number of small openings through which the evolved hydrocarbons pass. The inner chamber is cooled by cold air or water pipes to condense the hydrocarbon which is then run off to purifying-apparatus. In a modification, the retort is heated from the inside, the hydrocarbon being condensed in an outer case. Another form of retort consists of a narrow flat chamber, heated from one side and with a cooled condensing-chamber on the other; or two retorts may be used, with one condensing-chamber between them.

  11. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Brinjes, J F

    1867-03-12

    The invention relates to means for conveying the material through rotary retorts for distilling shale or the like. The retort is fitted with longitudinal ribs which lift the material and allow it to fall again as the retort rotates. Inclined deflecting plates attached to a fixed shaft cause the material as it falls to be gradually fed towards the discharge end of the retort. By means of the handle, which can be fixed in angular position by a pin entering holes in a quadrant, the angle of the plates may be adjusted and the rate of feed may be thus regulated. Or the plates may be hinged on the shaft or to the inside of the retort, and the angle is then adjusted by a longitudinal rod moved by a handwheel and nut. A similar arrangement may be applied to retorts with an oscillating motion.

  12. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Aitken, H; McAlley, R

    1871-05-18

    Shale or clay is coked or carbonized by the action of heat in retorts in the same way that coal is coked or by any of the methods used for coking or carbonizing coal, ironstone, or wood. Clay or shale, which is poor in carbon, is mixed or ground with coal, moss, peat, or earth mold, oil, tar, or other carbonaceous matter, shale, or the coke of certain kinds of coal after having been used in the manufacture of gas or oil by distillation. The mixture is coked or carbonized or the coke may be used alone and submitted to further coking or carbonization. The volatile hydrocarbons may be used in carbonizing or assisting to carbonize the shale by being burned beneath the retorts or they may be condensed along with sulfur, ammonia, etc., and kept for after use.

  13. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Greenshields, J

    1870-01-13

    In distilling shale or other carbonaceous substances to obtain oil, paraffin, ammonium salts, etc., sulfuric or other acid is added to the material, in the still or before treatment. The shale is first reduced to a fine powder. Acid tar from the treatment of oils may be used instead of acid. Hydrogen is sometimes passed into the still, or iron or other metal is mixed with the shale to generate hydrogen in the still. A figure is included which shows the condenser for paraffin and heavy oils, consisting of an iron cylinder connected with the still by a short pipe and surmounted by a long pipe communicating with other condensers. The pipe projects into the cylinder and perforated plates or baffles are fixed across the central portion of the cylinder. The condensed oils are drawn off by a pipe.

  14. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Walton, G

    1865-05-16

    A retort for the destructive distillation of coal, shale, whereby hydrocarbons are produced, is described. The vertical retort is provided with a charging door, a discharging door, an outlet leading to the condensing plant, an inclined bottom, and a perforated cage to facilitate the escape of the vapor and to regulate the amount of materials operated upon in the retort. The upper part of the cage is conical to deflect the materials fed in by the door and the lower part is also slightly conical to facilitate emptying the retort. The bottom may incline from both back and front, and also from the sides to the center. The apparatus is heated from below, and the flues pass all round the lower part of the retort.

  15. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

  16. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Young, W; Neilson, A; Young, A

    1876-10-09

    The invention relates to modifications of the retort apparatus, described in Specification No. 2487, A.D. 1872, for the destructive distillation of shale and other bituminous substances. The retorts instead of being worked continuously are completely filled and completely discharged in turn. They are made oblong in cross-section in order to present the material in thin layers and cause it to be acted upon more rapidly and economically. The retorts can thus be heated solely by the combustion of the carbonaceus matter contained in the discharged residues or with a small amount of coal in addition. Each retort is contracted at the bottom and is fitted with a box or chest having a hole in it corresponding to the opening in the retort and a sliding plate of iron, firebrick, or other suitable material, which can be operated by a rod passing through the front of the box, for opening or closing the retort. Underneath the box and over the combustion chamber are placed fireclay blocks leaving an opening, which can be closed by another plate of firebrick or the like. When distillation commences, the gases and vapors in the retort are drawn off through a pipe and a main by an exhauster. In order to prevent air from entering the retort or hydrocarbon vapor from being puffed back by the action of the wind, the gas which remains after the condensation of the oils is forced back into the box between the plates and part of it enters the retort and part the combustion chamber. In order to avoid the liability of the oil being carried past the condensers by the action of the gas, steam may be used as a substitute for the gas or mixed with it in large proportions, a steam jet being used to force the gas into the main supplying the boxes.

  17. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Young, W

    1875-09-07

    The invention relates to apparatus for absorbing and recovering volatile hydrocarbons from gases resulting from the distillation of coal, shale, and the like for the production of mineral oil, or from the coking of coal, charring of wood, and like operations. The apparatus is adapted for the repeated use of the same quantity of absorbent. The gases are passed by pipes through a coke tower down which is allowed to flow a stream of mineral oil or a fatty oil such as rape, olive, lard, tallow, and fish oils. The oil is supplied by a pipe and distributed by a plate which is held up against the opening of the pipe by a spring. The oil absorbs the volatile hydrocarbons contained in the gas and is drawn off by a pipe and passed through a tubular heat-exchanging apparatus, and thence by a pipe to a still formed at the bottom of the coke tower and constructed with a number of trays on the principle of the coffey still. The still is heated by a fire or by admitting the exhaust steam from an engine driving a fan, which draws the gases through the coke tower. The stream is admitted by a pipe and distributed by a perforated disk. Live steam may also be admitted if necessary by a pipe. The volatile oils are distilled off through a pipe to a coil condenser, and flow thence into a tank for separating any condensed water. This tank is covered preferably by a glass plate to allow inspection. The oil remaining in the still flows by a pipe to the heat-exchanger, in which it gives up heat to the oil flowing from the coke tower. It then flows by a pipe through a tubular cooler, cooled by water circulation, and thence to a store tank from which it is again pumped to the top of the coke tower.

  18. 30 CFR 36.50 - Tests of fuel tank.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Tests of fuel tank. 36.50 Section 36.50 Mineral... Requirements § 36.50 Tests of fuel tank. The fuel tank shall be inspected and tested to determine whether: (a) It is fuel-tight, (b) the vent maintains atmospheric pressure within the tank, and (c) the vent and...

  19. Irradiation test and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  20. Fuel motion in overpower tests of metallic integral fast reactor fuel

    International Nuclear Information System (INIS)

    Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

    1992-01-01

    In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

  1. 14 CFR 25.965 - Fuel tank tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 25.965 Section 25.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.965 Fuel tank tests. (a) It must be...

  2. Simulation of integral local tests with high-burnup fuel

    International Nuclear Information System (INIS)

    Gyori, G.

    2011-01-01

    The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)

  3. MCO Engineering Test Report Fuel Basket Handling Grapple Acceptance Test

    International Nuclear Information System (INIS)

    CHENAULT, D.M.

    2000-01-01

    Acceptance testing of the production SNF Fuel Basket lift grapples to the required 150 percent maximum lift load is documented herein. The report shows the results affirming the proof test passage. The primary objective of this test was to confirm the load rating of the grapple per applicable requirements of ANSI 14 6 American National Standard For Radioactive Materials Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500kg) or More. The above Standard requires a load test of 150% of the design load which must be held for a minimum of 10 minutes followed by a Liquid Penetrant or Magnetic Particle examination of critical areas and welds in accordance with the ANSI/ASME Boiler and Pressure Vessel Code 1989 Section 111 Division 1 section NF 5350

  4. Results of a conventional fuel tank blunt impact test

    Science.gov (United States)

    2015-03-23

    The Federal Railroad Administrations Office of Research : and Development is conducting research into passenger : locomotive fuel tank crashworthiness. A series of impact tests is : being conducted to measure fuel tank deformation under two : type...

  5. Distilling shale

    Energy Technology Data Exchange (ETDEWEB)

    Justice, P M

    1917-09-19

    Light paraffin oils and other oils for motors are obtained from shale, and benzene, toluene, and solvent naphtha are obtained from coal by a process in which the coal or shale is preferably powered to pass through a mesh of 64 to the inch and is heated with a mixture of finely ground carbonate or the like which under the action of heat gives off carbonic acid, and with small iron scrap or its equivalent which is adapted to increase the volume of hydrocarbons evolved. The temperature of the retort is maintained between 175 and 800/sup 0/C., and after all the vapors are given off at the higher temperature a fine jet of water may be injected into the retort and the temperature increased. The produced oil is condensed and purified by fractional distillation, and the gas formed is stored after passing it through a tower packed with coke saturated with a non-volatile oil with recovery of an oil of light specific gravity which is condensed in the tower. The residuum from the still in which the produced oil is fractionated may be treated with carbonate and iron, as in the first stage of the process, and the distillate therefrom passed to a second retort containing manganese dioxide and iron scrap preferably in the proportion of one part or two. The mixture, e.g., one containing shale or oil with six to thirteen percent of oxygen, to which is added three to eight per cent of carbonate, and from one and a half to four per cent of scrap iron, is conveyed by belts and an overhead skip to a hopper of a retort in a furnace heated by burners supplied with producer gas. The retort is fitted with a detachable lid and the vapors formed are led by a pipe to a vertical water-cooled condenser with a drain-cock which leads the condensed oils to a tank, from which a pipe leads to a packed tower for removing light oils and from which the gas passes to a holder.

  6. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Young, W; Fyfe, J

    1897-06-03

    Improvements in retorts of the class described in Specification No. 1377, A. D. 1882, for the destructive distillation of shale are disclosed. The retorts are provided with enlarged multiple hoppers for the reception of the fresh shale, and with enlarged chambers for the reception of the exhausted shale. The hoppers are built up of steel plates, and are bolted at the bottom to flanges on the upper ends of the retorts so as to permit of differential expansion. The shale is fed continuously into the retorts by rods or chains carried by a rocking shaft, or by a slit tube attached to a rocking shaft, and in connection with the hydraulic main. The spent shale is discharged into the receiving chambers by means of a series of prongs extending through a grating and carried by a rocking shaft actuated by levers engaging with reciprocating bars. In an alternative arrangement, the pronged rocking shafts are replaced by worms or screws formed into one half with a right-hand thread and the other half with a left-hand thread.

  7. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Bell, T

    1862-03-03

    Apparatus for the primary distillation of shale or other bituminous minerals in large quantity is constructed as follows:--An annular retort chamber is heated by two or more furnaces, one of which communicates with a central or internal vertical flue, and the others with external or encircling flues preferably disposed in zig-zag, helical or other tortuous course; or the gases pass up the external flues and down the internal flue or flues. The retort chamber may be divided by partitions, or there may be two or more separate chambers disposed concentrically or otherwise with intermediate flues. A pipe or pipes are provided to carry away volatile matters, and valved hoppers are arranged at the top of the retort chamber. The refuse or waste passes off by discharge tubes between the furnaces, and the mouths of these tubes dip into water tanks. The bottom of the retort chamber is funnel-shaped at the discharging points. The apparatus is preferably cylindrical, but may be triangular, square, or polygonal, and may be inclined or horizontal instead of vertical.

  8. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Brinjes, J F

    1866-03-13

    Continuous distillation of shale, boghead coal, or other noncaking bituminous material is effected in one or more rotating or oscillating horizontal retorts constructed to advance the charge. In the former case, internal spiral ribs or projections are provided. In the latter case, the retort is subdivided by annular flanges provided each with an opening, and a series of double inclined projections is disposed opposite the openings. An apparatus is shown in which the material is fed continuously from a hopper, and is advanced through upper and lower oscillating retorts, provided with annular flanges, double inclined projections, and longitudinal ribs or ledges. The retorts are supported on antifriction rollers. The retort is oscillated by means of a mangle wheel and a pinion on a shaft connected by a universal joint to a driving-shaft. The retort is oscillated from the retort by means of a chain connection. The retort is situated in a chamber separated by perforated brickwork from the actual furnace chamber, so that it is subjected to a lower temperature than the retort. The hopper delivers to crushing-rollers in a lower hopper which delivers to a shoot controlled by a sliding door. A hook on the retort is connected by a pipe to the retort, and a pipe leads from the hood to a condenser. A hood at the delivery end of the retort is connected by a pipe to an airtight cooler for the residue, which is discharged through doors into a truck of other receiver.

  9. Distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Craig, A

    1865-10-04

    To prevent oil distilled from coal, shale, or other minerals from being condensed and burnt in the retort, the oil is drawn off from the bottom of the retort. In order that the oil may be drawn off free from dirt and dust, the vertical retort is made of greater diameter at the bottom than at the top, and a vessel containing water is placed at the bottom. Within the retort is a cylinder built up of spaced rings, between which the oil percolates to the interior of the cylinder, whence it is drawn off through a pipe near its lower end. Externally, the rings present a smooth surface which offers no obstruction to the descent of the coal, and the passing of dust and dirt to the interior of the cylinder is prevented by making the lower edge of each ring overlap the upper edge of the ring below it. The cylinder may be replaced by a square, or other casing, and may be cast in one piece.

  10. Behavior of fission products released from severely damaged fuel during the PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hagrman, D.L.; Broughton, J.M.; Rest, J.

    1984-01-01

    The results of fission product release behavior during the first two Power Burst Facility Severe Fuel Damage tests are presented. Measured fission product release is compared with calculated release using temperature dependent release rate correlations and FASTGRASS analysis. The test results indicate that release from fuel of the high volatility fission products (Xe, Kr, I, Cs, and Te) is strongly influenced by parameters other than fuel temperature; namely fuel/fission product morphology, fuel and cladding oxidation state, extent of fuel liquefaction, and quench induced fuel shattering. Fission product transport from the test fuel through the sample system was strongly influenced by chemical effects. Holdup of I and Cs was affected by fission product chemistry, and transport time while Te release was primarily influenced by the extent of zircaloy oxidation. Analysis demonstrates that such integral test data can be used to confirm physical, chemical, and mechanistic models of fission product behavior for severe accident conditions

  11. Cofermentation of sweet sorghum juice and grain for production of fuel ethanol and distillers' wet grain

    Energy Technology Data Exchange (ETDEWEB)

    Gibbons, W.R.; Westby, C.A.

    1989-01-01

    In an attempt to reduce the costs associated with fuel ethanol production from grain, sweet sorghum juice was used as a partial or complete replacement for tap-water in mash preparation and fermentation. This juice, which was an unutilized by-product of sweet sorghum silage preservation by the Ag-Bag method, contained 6.5-7.6% (wt/wt) reducing sugar and produced up to 3.51% (v/v) ethanol beers after fermentation. Varying amounts of this juice were mixed with water and corn or wheat, either before or after liquefaction (front-end or back-end loading, respectively). When over 60% juice replacement was used in front-end loading trials, salt buildup, due to required pH adjustments during cooking, inhibited yeast metabolism and thereby reduced yields. This inhibition was not observed during back-end loading trials since acid and base usage during cooking were reduced. (author).

  12. Pre-irradiation testing of experimental fuel elements

    International Nuclear Information System (INIS)

    Basova, B.G.; Davydov, E.F.; Dvoretskij, V.G.; Ivanov, V.B.; Syuzev, V.N.; Timofeev, G.A.; Tsykanov, V.A.

    1979-01-01

    The problems of testing of experimental fuel elements of nuclear reactors on the basis of complex accountancy of the factors defining operating capacity of the fuel elements are considered. The classification of the parameters under control and the methods of initial technological testing, including testing of the fuel product, cladding and fished fuel element, is given. The requirements to the apparatus used for complex testing are formulated. One of the possible variants of representation of the information obtained in the form of the input certificate of a single fuel element under study is proposed. The processing flowsheet of the gathered information using the computer is given. The approach under consideration is a methodological basis of investigation of fuel element operating life at the testing stage of the experimental fuel elements

  13. 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.

  14. Feasibility study of the Dragon reactor for HTGR fuel testing

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

    The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

  15. Destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Hislop, G R

    1882-11-03

    Relates to apparatus for the treatment of cannel or common coal, shale, dross, peat, wood, and similar carbonaceous materials and ironstones for the purpose of obtaining gas, ammonia, and oil. A series of horizontal retorts are built into an arched chamber, and are supported by open arches. A series of vertical retorts in a chamber are situated beneath and in front of the retorts, so that the contents of the latter may easily be discharged into them. The carbonaceous material is first subjected to distillation in the retorts, the products passing by pipes to a hydraulic main where the coal tar and mineral oil are collected in the usual way. The gas is passed through oxide of iron and of lime if to be used for illuminating purposes, and through the former only, if to be used solely for heating purposes. The lower ends of the retorts are closed by doors, or may be sealed by water. They are preferably oblong in section and are surrounded by heating-flues, and each preferably contains the spent material from two of the primary retorts. They discharge their contents into a chamber at the bottom, from which they are withdrawn through a door. When the coke has been transferred from the horizontal to the vertical retorts the latter are closed by suitable covers, and the former are recharged with raw material. Superheated steam is introduced into the lower ends of the vertical retorts in order to facilitate the production of ammonia, which, together with the gases generated pass by a pipe to a main.

  16. Fast Flux Test Facility fuel and test management: The first 10 years

    International Nuclear Information System (INIS)

    Bennett, R.A.; Bennett, C.L.; Campbell, L.R.; Dobbin, K.D.; Tang, E.L.

    1991-07-01

    Core design and fuel and test management have been performed efficiently at the Fast Flux Test Facility. No outages have been extended to adjust core loadings. Development of mixed oxide fuels for advanced liquid metal breeder reactors has been carried out successfully. In fact, the fuel performance is extraordinary. Failures have been so infrequent that further development and refinement of fuel requirements seem appropriate and could lead to a significant reduction in projected electrical busbar costs. The Fast Flux Test Facility is also involved in early metal fuel development tests and appears to be an ideal test bed for any further fuel development or refinement testing. 3 refs., 4 figs., 2 tabs

  17. Posttest examination results of recent treat tests on metal fuel

    International Nuclear Information System (INIS)

    Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

    1986-01-01

    A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins

  18. Testing plutonium fuel assembly production for fast-neutron reactors

    International Nuclear Information System (INIS)

    Nougues, B.; Benhamou, A.; Bertothy, G.; Lepetit, H.

    1975-01-01

    The main characteristics of plutonium fuel elements for fast breeder reactors justify specific test procedures and special techniques. The specific tests relating to the Pu content consist of Pu enrichment and distribution tests, determination of the O/M ratio and external contamination tests. The specific tests performed on fuel configuration are: testing of sintered pellet diameter, testing of pin welding and checking of internal assmbly [fr

  19. Comprehensive Assessment of Composition and Thermochemical Variability by High Resolution GC/QToF-MS and the Advanced Distillation-Curve Method as a Basis of Comparison for Reference Fuel Development.

    Science.gov (United States)

    Lovestead, Tara M; Burger, Jessica L; Schneider, Nico; Bruno, Thomas J

    2016-12-15

    Commercial and military aviation is faced with challenges that include high fuel costs, undesirable emissions, and supply chain insecurity that result from the reliance on petroleum-based feedstocks. The development of alternative gas turbine fuels from renewable resources will likely be part of addressing these issues. The United States has established a target for one billion gallons of renewable fuels to enter the supply chain by 2018. These alternative fuels will have to be very similar in properties, chemistry, and composition to existing fuels. To further this goal, the National Jet Fuel Combustion Program (a collaboration of multiple U.S. agencies under the auspices of the Federal Aviation Administration, FAA) is coordinating measurements on three reference gas turbine fuels to be used as a basis of comparison. These fuels are reference fuels with certain properties that are at the limits of experience. These fuels include a low viscosity, low flash point, high hydrogen content "best case" JP-8 (POSF 10264) fuel, a relatively high viscosity, high flash point, low hydrogen content "worst case" JP-5 (POSF 10259) fuel, and a Jet-A (POSF 10325) fuel with relatively average properties. A comprehensive speciation of these fuels is provided in this paper by use of high resolution gas chromatography/quadrupole time-of-flight - mass spectrometry (GC/QToF-MS), which affords unprecedented resolution and exact molecular formula capabilities. The volatility information as derived from the measurement of the advanced distillation curve temperatures, T k and T h , provides an approximation of the vapor liquid equilibrium and examination of the composition channels provides detailed insight into thermochemical data. A comprehensive understanding of the compositional and thermophysical data of gas turbine fuels is required not only for comparison but also for modeling of such complex mixtures, which will, in turn, aid in the development of new fuels with the goals of

  20. Comprehensive Assessment of Composition and Thermochemical Variability by High Resolution GC/QToF-MS and the Advanced Distillation-Curve Method as a Basis of Comparison for Reference Fuel Development*

    Science.gov (United States)

    Lovestead, Tara M.; Burger, Jessica L.; Schneider, Nico; Bruno, Thomas J.

    2018-01-01

    Commercial and military aviation is faced with challenges that include high fuel costs, undesirable emissions, and supply chain insecurity that result from the reliance on petroleum-based feedstocks. The development of alternative gas turbine fuels from renewable resources will likely be part of addressing these issues. The United States has established a target for one billion gallons of renewable fuels to enter the supply chain by 2018. These alternative fuels will have to be very similar in properties, chemistry, and composition to existing fuels. To further this goal, the National Jet Fuel Combustion Program (a collaboration of multiple U.S. agencies under the auspices of the Federal Aviation Administration, FAA) is coordinating measurements on three reference gas turbine fuels to be used as a basis of comparison. These fuels are reference fuels with certain properties that are at the limits of experience. These fuels include a low viscosity, low flash point, high hydrogen content “best case” JP-8 (POSF 10264) fuel, a relatively high viscosity, high flash point, low hydrogen content “worst case” JP-5 (POSF 10259) fuel, and a Jet-A (POSF 10325) fuel with relatively average properties. A comprehensive speciation of these fuels is provided in this paper by use of high resolution gas chromatography/quadrupole time-of-flight – mass spectrometry (GC/QToF-MS), which affords unprecedented resolution and exact molecular formula capabilities. The volatility information as derived from the measurement of the advanced distillation curve temperatures, Tk and Th, provides an approximation of the vapor liquid equilibrium and examination of the composition channels provides detailed insight into thermochemical data. A comprehensive understanding of the compositional and thermophysical data of gas turbine fuels is required not only for comparison but also for modeling of such complex mixtures, which will, in turn, aid in the development of new fuels with the goals of

  1. Interim report spent nuclear fuel retrieval system fuel handling development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  2. 14 CFR 23.965 - Fuel tank tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 23.965 Section 23.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.965 Fuel...

  3. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  4. A comprehensive in-pile test of PWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

  5. Performance indicators of bioethanol distillation

    International Nuclear Information System (INIS)

    Marriaga, Nilson

    2009-01-01

    The increase of biofuels demand accelerates the construction of new production plants and technological improvements in the process so the development of versatile tools for evaluating alternatives becomes an undeniable challenge. It was established through heuristic rules, thermodynamic analysis and simulation computer the energy consumption and performance indicators that govern, from fermented mash (ethanol 8.5 % v/v), the distillation of various capacities for bioethanol production: 20, 60, 100 and 150 KLD (kiloliters / day) through Aspen PlusTM simulator. It was found that the distillation demand nearly 30% of heat that would be obtained by burning alcohol fuel produced thus it is necessary the use of raw materials that generate enough biomass to produce the steam required. In addition, correlations were found to allow for easy diameters of distillation columns in terms of production capacity.

  6. Sodium flow distribution in test fuel assembly P-23B

    International Nuclear Information System (INIS)

    Taylor, J.P.S.

    1978-08-01

    Relatively large cladding diametral increases in the exterior fuel pins of HEDL's test fuel subassembly P-23B were successfully explained by a thermal-hydraulic/solid mechanics analysis. This analysis indicates that while at power, the subassembly flow was less than planned and that the fuel pins were considerably displaced and bowed from their nominal position. In accomplishing this analysis, a method was developed to estimate the sodium flow distribution and pin distortions in a fuel subassembly at power

  7. Test reports for K Basins vertical fuel handling tools

    Energy Technology Data Exchange (ETDEWEB)

    Meling, T.A.

    1995-02-01

    The vertical fuel handling tools, for moving N Reactor fuel elements, were tested in the 305 Building Cold Test Facility (CTF) in the 300 Area. After fabrication was complete, the tools were functionally tested in the CTF using simulated N Reactor fuel rods (inner and outer elements). The tools were successful in picking up the simulated N Reactor fuel rods. These tools were also load tested using a 62 pound dummy to test the structural integrity of each assembly. The tools passed each of these tests, based on the performance objectives. Finally, the tools were subjected to an operations acceptance test where K Basins Operations personnel operated the tool to determine its durability and usefulness. Operations personnel were satisfied with the tools. Identified open items included the absence of a float during testing, and documentation required prior to actual use of the tools in the 100 K fuel storage basin.

  8. Destructive distillation of coals

    Energy Technology Data Exchange (ETDEWEB)

    Rollason, A

    1918-08-23

    To obtain light oils and ammonia from coals having volatile and oxygen contents, the crushed material is mixed with 5 percent of ground amorphous calcium carbonate and distilled slowly in a cast iron retort to remove the water and light oils, the ammonia being synthesized at a later stage. The crushed residue is gasified in a producer by a blast of air and superheated steam at about 950/sup 0/C. The steam and air are passed very slowly at low pressure through the fuel to cause the dissociation of the atmospheric nitrogen molecules into atoms. The gases are then passed to a heater, having a temperature of 500/sup 0/C, and thence to a continuously working externally-heated retort charged with fuel, such as the hard retort residues, maintained below 850/sup 0/C. The water vapor in the gases is dissociated by the incandescent fuel, the oxygen combining with the carbon, and the lime present in the fuel causes the hydrogen to combine with the free nitrogen atoms, thus forming ammonia. The gases after leaving the retort are cooled down to 85 to 95/sup 0/C and the ammonia may be recovered by conversion into ammonium sulphate. The resultant cooled gases may again be charged with superheated steam and utilized again in the heater and retort.

  9. Test plan for K-Basin fuel handling tools

    International Nuclear Information System (INIS)

    Bridges, A.E.

    1995-01-01

    The purpose of this document is to provide the test plan and procedures for the acceptance testing of the handling tools enveloped for the removal of an N-Reactor fuel element from its storage canister in the K-Basins storage pool and insertion into the Single fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N-Reactor fuel elements is part of the overall characterization effort. New hand tools were required since previous fuel movement has involved grasping the fuel in a horizontal position. The 305 Building Cold Test Facility will be used to conduct the acceptance testing of the Fuel Handling Tools. Upon completion of this acceptance testing and any subsequent training of operators, the tools will be transferred to the 105 KW Basin for installation and use

  10. Structural geology report: Spent Fuel Test - Climax Nevada Test Site

    International Nuclear Information System (INIS)

    Wilder, D.G.; Yow, J.L. Jr.

    1984-10-01

    We performed underground mapping and core logging in the Climax Stock, a granitic intrusive at the Nevada Test Site, as part of a major field test to determine the feasibility of using granitic or crystalline rock for the underground storage of spent fuel from a nuclear reactor. This mapping and logging identified more than 2500 fractures, over 1500 of which were described in enough detail to allow statistical analyses and orientation studies to be performed. We identified eight joint sets, three major shear sets, and a fault zone within the Spent Fuel Test - Climax (SFT-C) portion of the Stock. Joint sets identified within the SFT-C and elsewhere in the Stock correlated well. The orientations of joint sets identified by other investigators were consistent with our findings, indicating that the joint sets are persistent and have a relatively uniform orientation throughout a major portion of the Stock. The one joint set not seen elsewhere in the Stock is healed and the wall rock is altered, implying that healed joints were not included in the mapping criteria used by other investigators. The shear sets were distinguished from the joint sets by virtue of crushed minerals, continuous clay infilling, and other evidences of shearing, and from faults by the lack of offsetting. Previous investigators working mainly in the Pile Driver Drifts identified two of the shear sets. The third set, being nearly parallel to these Drifts had not been identified previously. The fault zone identified at the far (Receiving Room) end of the project is oriented approximately N45 0 E-75 0 SE, similar to both the Boundary and Shaft Station Faults. We have, therefore, concluded that the Receiving Room Fault is one of a series of normal faults that occur within the Climax Stock and that are possibly related, in both age and genesis, to the Boundary Fault. 52 refs., 26 figs., 11 tabs

  11. Interim results from UO2 fuel oxidation tests in air

    International Nuclear Information System (INIS)

    Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

    1987-08-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

  12. LEU WWR-M2 fuel assemblies burnable test

    International Nuclear Information System (INIS)

    Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

    2000-01-01

    The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

  13. Distilling solid carbonaceous materials

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, H; Laing, B

    1926-12-04

    In a process of distilling solid carbonaceous materials with by-product recovery, the time factor and the temperature gradient during the distillation period are so controlled that a temperature difference exceeding 150/sup 0/C is avoided between the temperatures at the center and periphery of any suitable size of material or thickness of fuel bed. The material is heated by direct contact with an inert gas, such as water gas, producer gas, or combustion gases, which is passed in counterflow to the material and whose volume is such as to lower the vapor tension or partial pressure of the volatilizable oils and to withdraw the oils without cracking of the oil vapors. The material may be subjected to a preliminary heat treatment by gases containing 2 to 3 percent of free oxygen to reduce its coking properties, and free oxygen may be added either to the heating gases during the heat treatment, or to the retort and heating gases and vapors to polymerize resinous bodies prior to condensation or during condensation and while the oils are still wholly or partially in the vapor state.

  14. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  15. Results of tests with open fuel in KNK II

    International Nuclear Information System (INIS)

    Schmitz, G.

    1987-03-01

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

  16. Fuels for research and test reactors, status review: July 1982

    International Nuclear Information System (INIS)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO 2 rod fuels. Among new fuels, those given major emphasis include H 3 Si-Al dispersion and UO 2 caramel plate fuels

  17. Distilling hydrocarbon oils

    Energy Technology Data Exchange (ETDEWEB)

    Tausz, J

    1924-07-16

    Hydrocarbon oils such as petroleum, shale oils, lignite or coal tar oils are purified by distilling them and collecting the distillate in fractions within narrow limits so that all the impurities are contained in one or more of the narrow fractions. In distilling ligroin obtained by destructive distillation of brown coal, it is found that the coloring and resin-forming constituents are contained in the fractions distilling over at 62 to 86/sup 0/C and 108/sup 0/C. The ligroin is purified, therefore, by distillating in an apparatus provided with an efficient dephlegmotor and removing these two fractions. The distillation may be carried out wholly or in part under reduced pressure, and fractions separated under ordinary pressure may be subsequently distilled under reduced pressure. The hydrocarbons may be first separated into fractions over wider limits and the separate fractions be subjected to a further fractional distillation.

  18. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    International Nuclear Information System (INIS)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

  19. Characteristic test technology for PWR fuel and its components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

  20. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  1. Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Creer, J.M.

    1988-10-01

    REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

  2. Current and prospective fuel test programmes in the MIR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

    2007-07-01

    MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

  3. Behavior of metallic fuel in treat transient overpower tests

    International Nuclear Information System (INIS)

    Bauer, T.H.; Wright, A.E.; Robinson, W.R.; Klickman, A.E.

    1988-01-01

    Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types

  4. Certification test for safety of new fuel transportation package

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Sugawa, Osami; Suga, Masao.

    1993-01-01

    The objective of this certification test is to prove the safety of new fuel transportation package against a fire of actual size caused by traffic accidents. After the fire test, the fuel assemblies were covered with coal-tar like material vaporized from anti-shock material used in the container. Surface color of BWR-type fuel assembly was dark grey that is supposed to be the color of oxide of Zircaloy. As for PWR-type fuel assembly, the condition encountered during fire test caused no change to the outlook of the rod element. Both the BWR and PWR type fuel rod elements showed no deformation and were completely sound. Therefore it may be concluded that the container protected the mimic fuel assemblies against fire of 30 minutes duration and caused no damage. This report is the result of the above experiments and examinations, and we appreciate the cooperation of those who are concerned. (J.P.N.)

  5. The Use of Fuel Gas as Stripping Medium in Atmospheric Distillation of Crude Oil L’utilisation de gaz combustible comme moyen d’extraction des fractions légères en distillation atmosphérique du pétrole brut

    OpenAIRE

    Plellis-Tsaltakis C.; Lygeros A.I.

    2011-01-01

    Stripping of petroleum fractions aims to remove the light ends that spoil some of their properties, such as the flash point. Stripping usually employs steam for that purpose. Except for steam, other substances can perform the same function, among them light hydrocarbons. In this article, we investigate the use of refinery fuel gas as stripping medium for a crude oil atmospheric distillation unit. L’extraction des fractions légères du pétrole vise à enlever les produits nuisibles à cer...

  6. Materials testing for molten carbonate fuel cells

    International Nuclear Information System (INIS)

    Di Mario, F.; Frangini, S.

    1995-01-01

    Unlike conventional generation systems fuel cells use an electrochemical reaction between a fossil fuel and an oxidant to produce electricity through a flame less combustion process. As a result, fuel cells offer interesting technical and operating advantages in terms of conversion efficiencies and environmental benefits due to very low pollutant emissions. Among the different kinds of fuel cells the molten carbonate fuel cells are currently being developed for building compact power generation plants to serve mainly in congested urban areas in virtue of their higher efficiency capabilities at either partial and full loads, good response to power peak loads, fuel flexibility, modularity and, potentially, cost-effectiveness. Starting from an analysis of the most important degradative aspects of the corrosion of the separator plate, the main purpose of this communication is to present the state of the technology in the field of corrosion control of the separator plate in order to extend the useful lifetime of the construction materials to the project goal of 40,000 hours

  7. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

  8. The 3rd irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Park, J. H. and others

    2001-05-01

    The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

  9. 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...

  10. 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)

  11. Design criteria and fabrication in-pile test section of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    Safety state fuel test loop will be equipped in HANARO to obtain the development and betterments of advanced fuel and materials through the irradiation tests. The objective of this study is to determine the design criteria and technical specification of in-pile test section and to specify the manufacturing requirements of in-pile test section. HANARO fuel test loop was designed to meet the CANDU and PWR fuel testing and in-pile section will be manufactured and installed in HANARO. The design criteria and technical specification of in-pile test section could be used the fuel and materials design with for irradiation testing IPS of HANARO fuel test loop. This results will become guidances for the planning and programming of irradiation testing. (author). 12 refs., tabs., figs.

  12. Drop testing of the Westinghouse fresh nuclear fuel package

    International Nuclear Information System (INIS)

    Shappert, L.B.; Sanders, C.F.

    1993-01-01

    The Westinghouse Columbia Fuel Fabrication Facility has decided to develop and certify a new fresh fuel package design (type A, fissile) that has the capability to transport more highly enriched fuel than was previously possible. A prototype package was tested in support of the Safety Analysis Report of the Packaging (SARP). This paper provides detailed information on the tests and test results. A first prototype test was carried out at the STF, and the design did not give the safety margin that Westinghouse wanted for their containers. The data from the test were used to redesign the connection between the clamping frame and the pressure pad, and the tests were reinitiated. Three packages were then tested at the STF. All packages met the acceptance criteria and acceleration information was obtained that provided an indication of the behavior of the cradle and strongback which holds the fuel assemblies and nuclear poison in place. (J.P.N.)

  13. Methodology of thermalhydraulic tests of fuel assemblies for WWER-1000

    International Nuclear Information System (INIS)

    Archipov, A.; Kolochko, V.N.

    2001-01-01

    At present 11 units with WWER-1000 are in operation in Ukraine. The NPPs are provided with nuclear fuel from Russia. The fuel assemblies are fabricated and delivered to Ukrainian NPPs from Russia. However the contemporary tendencies of nuclear energy development in the world assume a diversification of nuclear fuel vendors. Therefore the creation of the own nuclear fuel cycle of Ukraine is in mind in the strategy of nuclear energy development of Ukraine. As a part of the fuel assemblies fabrication process complex of the thermalhydraulic tests should be carried out to confirm design characteristics of the fuel assemblies before they are loaded in the reactor facility. The experimental basis and scientific infrastructure for the thermalhydraulic tests arrangement and realization of the programs and procedures for the core equipment examination are under consideration. (author)

  14. Conceptual development of a test facility for spent fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

  15. Conceptual development of a test facility for spent fuel management

    International Nuclear Information System (INIS)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

    1997-01-01

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

  16. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  17. Fabrication of Fast Reactor Fuel Pins for Test Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

    1967-09-15

    An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)

  18. Catalytic distillation structure

    Science.gov (United States)

    Smith, L.A. Jr.

    1984-04-17

    Catalytic distillation structure is described for use in reaction distillation columns, and provides reaction sites and distillation structure consisting of a catalyst component and a resilient component intimately associated therewith. The resilient component has at least about 70 volume % open space and is present with the catalyst component in an amount such that the catalytic distillation structure consists of at least 10 volume % open space. 10 figs.

  19. In-pile tests of HTGR fuel particles and fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

    1985-01-01

    Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

  20. Characterization of spent fuel approved testing material---ATM-105

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  1. Final report of fuel dynamics Test E7

    International Nuclear Information System (INIS)

    Doerner, R.C.; Murphy, W.F.; Stanford, G.S.; Froehle, P.H.

    1977-04-01

    Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release [there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow]; (3) the predominant postfailure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4) less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets

  2. Operational reliability testing of FBR fuel in EBR-II

    International Nuclear Information System (INIS)

    Asaga, Takeo; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

    1991-01-01

    The operational reliability testing of FBR fuel has been conducting in EBR-II as a DOE/PNC collaboration program. This paper reviews the achieved summary of Phase-I test as well as outline of progressing Phase-II test. In Phase-I test, the reliability of FBR fuel pins including 'MONJU' fuel was demonstrated at the event of operational transient. Continued operation of the failed pins was also shown to be feasible without affecting the plant operation. The objectives of the Phase-II test is to extend the data base relating with the operational reliability for long life fuel, and to supply the highly quantitative evaluation. The valuable insight obtained in Phase-II test are considerably expected to be useful toward the achievement of commercial FBR. (author)

  3. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

    1988-05-01

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  4. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  5. Furnaces for the distillation of coal, etc

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, F A; Deacon, M; Brady, N P.W.

    1918-08-14

    A tunnel or other furnace for the distillation of coal of the kind provided with inverted pockets in its roof to collect diverse distillates in the manner described, characterized by one or more of the pockets being provided with a sloping roof whose gradient from the higher end downwards is in the direction of the forward travel of the fuel beneath it for the purposes described.

  6. Remote helium leak test of the DUPIC fuel rod

    International Nuclear Information System (INIS)

    Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

    1998-01-01

    DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

  7. Testing Systems and Results for Advanced Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

    2012-01-01

    Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

  8. Tomographic imaging of severely disrupted fuel assemblies tested in TREAT

    International Nuclear Information System (INIS)

    Morman, J.A.; Froehle, P.H.; Holland, J.W.; Bennett, J.D.

    1990-01-01

    A series of CT codes is under development in the Reactor Analysis and Safety Division of Argonne National Laboratory for use as a post-test examination tool to analyze segments of the final fuel-bundle configuration of TREAT tests. This paper presents the results of CT analysis for fuel assemblies using neutron radiography. Fuel relocation following overpower transients in the TREAT reactor is examined for sections of the assemblies, and results are compared to metallographic sections. Further improvements are expected to increase the use and reliability of CT analysis as a standard post-test examination tool

  9. Pie technique of LWR fuel cladding fracture toughness test

    International Nuclear Information System (INIS)

    Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu

    2006-01-01

    Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)

  10. Test of fuel handling machine for Monju in sodium

    International Nuclear Information System (INIS)

    Ishii, Yoichiro; Masuda, Yoichi; Kataoka, Hajime

    1980-01-01

    Various types of fuel handling machines were studied, and under-the-plug method of fuel exchange and the fuel handling machine of single turning plug, fixed arm type were selected for the prototype reactor ''Monju'', because the turning plug is relatively small, and the rate of operation, safety, operational ability, maintainability and reliability required for the reactor are satisfied, moreover, the extrapolation to the demonstration reactor was considered. Attention must be paid to the points that the fuel handling machine is very long and invisible from outside, and the smooth operation and endurance in sodium are required for it. The full mock-up testing facility of single turning plug, fixed arm type was installed in 1974, and the full mock-up test has been carried out since 1975 in Oarai. Fuel exchange is carried out at about 6 months intervals in Monju, and about 20 to 30% of core and blanket fuels are exchanged for about one month period. The functions required for the fuel handling machine for Monju, the outline of the testing facility, the schedule of the testing, the items of testing and the results, and the matters to be specially written are described. The full mock-up test in sodium has been carried out for 5 years, and the functions and the endurance have been proved sufficiently. (Kako, I.)

  11. Distillation of shale in situ

    Energy Technology Data Exchange (ETDEWEB)

    de Ganahl, C F

    1922-07-04

    To distill buried shale or other carbon containing compounds in situ, a portion of the shale bed is rendered permeable to gases, and the temperature is raised to the point of distillation. An area in a shale bed is shattered by explosives, so that it is in a relatively finely divided form, and the tunnel is then blocked by a wall, and fuel and air are admitted through pipes until the temperature of the shale is raised to such a point that a portion of the released hydrocarbons will burn. When distillation of the shattered area takes place and the lighter products pass upwardly through uptakes to condensers and scrubbers, liquid oil passes to a tank and gas to a gasometer while heavy unvaporized products in the distillation zone collect in a drain, flow into a sump, and are drawn off through a pipe to a storage tank. In two modifications, methods of working are set out in cases where the shale lies beneath a substantially level surface.

  12. Spent fuel handling system for a geologic storage test at the Nevada Test Site

    International Nuclear Information System (INIS)

    Duncan, J.E.; House, P.A.; Wright, G.W.

    1980-01-01

    The Lawrence Livermore Laboratory is conducting a test of the geologic storage of encapsulated spent commercial reactor fuel assemblies in a granitic rock at the Nevada Test Site. The test, known as the Spent Fuel Test-Climax (SFT-C), is sponsored by the US Department of Energy, Nevada Operations Office. Eleven pressurized-water-reactor spent fuel assemblies are stored retrievably for three to five years in a linear array in the Climax stock at a depth of 420 m

  13. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  14. Characterization of spent fuel approved testing material: ATM-106

    International Nuclear Information System (INIS)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

    1988-10-01

    The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs

  15. Spent fuel drying system test results (second dry-run)

    International Nuclear Information System (INIS)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks have been detected in the basins and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the second dry-run test, which was conducted without a fuel element. With the concurrence of project management, the test protocol for this run, and subsequent drying test runs, was modified. These modifications were made to allow for improved data correlation with drying procedures proposed under the IPS. Details of these modifications are discussed in Section 3.0

  16. 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...

  17. The role of spent fuel test facilities in the fuel cycle strategy

    International Nuclear Information System (INIS)

    Huang, S. T.; Gross, D. L.; Snyder, N. W.; Woods, W. D.

    1988-01-01

    Disposal of commercial spent nuclear fuels in the major industrialized countries may be categorized into two broad approaches: a once-through policy which will dispose of spent fuels and recycle fissile materials. Within reprocess spent fuels and recycle fissile materials. Within each policy, various technical, licensing, institutional and public issues exist. These issues tend to complicate the formulation of an effective and acceptable fuel cycle strategy which will meet various cost, schedule, and legislative constraints. This paper examines overall fuel cycle strategies from the viewpoint of these underlying technical issues and assesses the roles of spent fuel test facilities in the overall fuel cycles steps. Basic functions of such test facilities are also discussed. The main emphasis is placed on the once-through policy although the reprocessing / recycle policy is also discussed. Benefits of utilizing test facilities in the fuel cycle strategies are explored. The results indicate that substantial benefits may be obtained in terms of minimizing programmatic risks, increasing public confidence, and more effective utilization of overall budgetary resources by structuring and highlighting the test facilities as an important element in the overall strategy

  18. Situation of test and research reactors' spent fuels

    International Nuclear Information System (INIS)

    Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

    1996-01-01

    The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

  19. Simulation and Test of a Fuel Cell Hybrid Golf Cart

    Directory of Open Access Journals (Sweden)

    Jingming Liang

    2014-01-01

    Full Text Available This paper establishes the simulation model of fuel cell hybrid golf cart (FCHGC, which applies the non-GUI mode of the Advanced Vehicle Simulator (ADVISOR and the genetic algorithm (GA to optimize it. Simulation of the objective function is composed of fuel consumption and vehicle dynamic performance; the variables are the fuel cell stack power sizes and the battery numbers. By means of simulation, the optimal parameters of vehicle power unit, fuel cell stack, and battery pack are worked out. On this basis, GUI mode of ADVISOR is used to select the rated power of vehicle motor. In line with simulation parameters, an electrical golf cart is refitted by adding a 2 kW hydrogen air proton exchange membrane fuel cell (PEMFC stack system and test the FCHGC. The result shows that the simulation data is effective but it needs improving compared with that of the real cart test.

  20. Metallographic analysis of irradiated RERTR-3 fuel test specimens

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-01-01

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

  1. Drop testing of the Westinghouse fresh nuclear fuel package

    International Nuclear Information System (INIS)

    Shappert, L.B.; Sanders, C.F.

    1992-01-01

    In recent years, the Westinghouse Columbia Fuel Fabrication Facility has been faced with increasing pressure from utilities that wished to take the fuel in their nuclear power plants to higher burnups. To help accommodate this trend, Westinghouse has determined that it needs the ability to increase the enrichment of the fresh fuel it delivers to its customers. One critical step in this process is to certify a new (Type A, fissile) fresh fuel package design that has the capability to transport fuel with a higher enrichment than was previously available. A prototype package was tested in support of the Safety Analysis Report of the Packaging. This paper provides detailed information on those tests and their results

  2. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  3. Feasibility study on the transient fuel test loop installation

    International Nuclear Information System (INIS)

    Kim, J. Y.; Lee, C. Y.

    1997-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. The objective of this study is to investigate and analyze the test capsules and loops in research reactors of the other countries and to design preliminarily the eligible transient fuel test facility to be installed in HANARO. The principle subjects of this study are to analyze the contents, kinds and scopes of the irradiation test facilities for nuclear technology development. The guidances for the basic and detail design of the transient fuel test facility in the future are presented. The investigation and analysis of various kinds of test facilities that are now in operation at the research reactors of nuclear advanced countries are carried out. Based on the design data of HANARO the design materials for an eligible transient fuel test facility comprises two pacts : namely, in pile test fuel in reactor core site, and out of pile system regulates the experimental conditions in the in pile test section. Especially for power ramping and cycling selection of the eligible power variation equipment in HANARO is carried out. (author). 13 refs., 4 tabs., 46 figs

  4. The Design and Manufacturing of Essential oil Distillation Plant for ...

    African Journals Online (AJOL)

    The Design and Manufacturing of Essential oil Distillation Plant for rural poverty ... The adaptation of oil distillation technology for essential oil production is ... based on local resources and the first prototype has been manufactured and tested.

  5. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  6. A High Rated Solar Water Distillation Unit for Solar Homes

    Directory of Open Access Journals (Sweden)

    Abhishek Saxena

    2016-01-01

    Full Text Available India is presently focusing on complete utilization of solar energy and saving fossil fuels, which are limited. Various solar energy systems like solar cookers, solar water heaters, solar lanterns, solar PV lights, and solar lamps are continuously availing by the people of India at a low cost and on good subsidies. Apart from this, India is a solar energy promising country with a good number of solar homes (carrying solar energy systems in its various locations. The present paper focuses on a unique combination of solar dish cooker (SDC and solar water heater (SWH to produce distilled water with a high distillate and a high daily productivity. The procedure has been discussed on the basis of experimental testing to produce distilled water by combining an evacuated type SWH and a SDC. Experimentation has been carried out in MIT, Moradabad (longitude, 28.83°N, and latitude, 78.78°E by developing the same experimental setup on behalf of solar homes. The daily productivity of distilled water was found around 3.66 litres per day in full sunshine hours for an approximated pH value of 7.7 and a ppm value of 21. The payback period (PBP has been estimated around 1.16 years of the present system.

  7. Driving Force Based Design of Cyclic Distillation

    DEFF Research Database (Denmark)

    Nielsen, Rasmus Fjordbak; Huusom, Jakob Kjøbsted; Abildskov, Jens

    2017-01-01

    with mixed phase feeds. A range of binary test cases, benzene toluene, methanol water, and ethanol water, are evaluated. The advantage of the design approach in cyclic distillation is shown to be analogous to the advantages obtained in conventional continuous distillation, including a minimal utility...

  8. Diesel fuel lubricity testing revisited : Tests von Dieselkraftstoffschmierfähigkeit erneut betrachtet

    NARCIS (Netherlands)

    van Leeuwen, H.J.

    2017-01-01

    Fuel is used as a lubricant in several engine components. Diesel fuel is known for its good lubrication properties, better than gasoline. These properties are examined in standard tests, as prescribed by ASTM. Good lubrication properties are designated as a good lubricity. Most commonly, fuel

  9. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

  10. Fuel performance analysis for the HAMP-1 mini plate test

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

  11. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  12. Radiochemical analyses of several spent fuel Approved Testing Materials

    International Nuclear Information System (INIS)

    Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

    1994-09-01

    Radiochemical characterization data are described for UO 2 and UO 2 plus 3 wt% Gd 2 O 3 commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, 14 C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program

  13. Results of industrial tests of carbonate additive to fuel oil

    Science.gov (United States)

    Zvereva, E. R.; Dmitriev, A. V.; Shageev, M. F.; Akhmetvalieva, G. R.

    2017-08-01

    Fuel oil plays an important role in the energy balance of our country. The quality of fuel oil significantly affects the conditions of its transport, storage, and combustion; release of contaminants to atmosphere; and the operation of main and auxiliary facilities of HPPs. According to the Energy Strategy of Russia for the Period until 2030, the oil-refining ratio gradually increases; as a result, the fraction of straight-run fuel oil in heavy fuel oils consistently decreases, which leads to the worsening of performance characteristics of fuel oil. Consequently, the problem of the increase in the quality of residual fuel oil is quite topical. In this paper, it is suggested to treat fuel oil by additives during its combustion, which would provide the improvement of ecological and economic indicators of oil-fired HPPs. Advantages of this method include simplicity of implementation, low energy and capital expenses, and the possibility to use production waste as additives. In the paper, the results are presented of industrial tests of the combustion of fuel oil with the additive of dewatered carbonate sludge, which is formed during coagulation and lime treatment of environmental waters on HPPs. The design of a volume delivery device is developed for the steady additive input to the boiler air duct. The values are given for the main parameters of the condition of a TGM-84B boiler plant. The mechanism of action of dewatered carbonate sludge on sulfur oxides, which are formed during fuel oil combustion, is considered. Results of industrial tests indicate the decrease in the mass fraction of discharged sulfur oxides by 36.5%. Evaluation of the prevented damage from sulfur oxide discharged into atmospheric air shows that the combustion of the fuel oil of 100 brand using carbonate sludge as an additive (0.1 wt %) saves nearly 6 million rubles a year during environmental actions at the consumption of fuel oil of 138240 t/year.

  14. Spent fuel drying system test results (first dry-run)

    International Nuclear Information System (INIS)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first dry-run test, which was conducted without a fuel element. The empty test apparatus was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The data from this dry-run test can serve as a baseline for the first two fuel element tests, 1990 (Run 1) and 3128W (Run 2). The purpose of this dry-run was to establish the background levels of hydrogen in the system, and the hydrogen generation and release characteristics attributable to the test system without a fuel element present. This test also serves to establish the background levels of water in the system and the water release characteristics. The system used for the drying test series was the Whole Element Furnace Testing System, described in Section 2.0, which is located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in section 3.0, and the experimental

  15. The US Army Foreign Comparative Test fuel cell program

    Science.gov (United States)

    Bostic, Elizabeth; Sifer, Nicholas; Bolton, Christopher; Ritter, Uli; Dubois, Terry

    The US Army RDECOM initiated a Foreign Comparative Test (FCT) Program to acquire lightweight, high-energy dense fuel cell systems from across the globe for evaluation as portable power sources in military applications. Five foreign companies, including NovArs, Smart Fuel Cell, Intelligent Energy, Ballard Power Systems, and Hydrogenics, Inc., were awarded competitive contracts under the RDECOM effort. This paper will report on the status of the program as well as the experimental results obtained from one of the units. The US Army has interests in evaluating and deploying a variety of fuel cell systems, where these systems show added value when compared to current power sources in use. For low-power applications, fuel cells utilizing high-energy dense fuels offer significant weight savings over current battery technologies. This helps reduce the load a solider must carry for longer missions. For high-power applications, the low operating signatures (acoustic and thermal) of fuel cell systems make them ideal power generators in stealth operations. Recent testing has been completed on the Smart Fuel Cell A25 system that was procured through the FCT program. The "A-25" is a direct methanol fuel cell hybrid and was evaluated as a potential candidate for soldier and sensor power applications.

  16. Concepts for Small-Scale Testing of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven Craig [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

  17. Power Burst Facility Severe Fuel Damage test series

    International Nuclear Information System (INIS)

    Buescher, B.J.; Osetek, D.J.; Ploger, S.A.

    1982-01-01

    The Severe Fuel Damage (SFD) tests planned for the Power Burst Facility (PBF) are described. Bundles containing 32 zircaloy-clad, PWR-type fuel rods will be subjected to severe overheating transients in a high-pressure, superheated-steam environment. Cladding temperatures are expected to reach 2400 0 K, resulting in cladding ballooning and rupture, severe cladding oxidation, cladding melting, fuel dissolution, fuel rod fragmentation, and possibly, rubble bed formation. An experiment effluent collection system is being installed and the PBF fission product monitoring system is being upgraded to meet the special requirements of the SFD tests. Scoping calculations were performed to evaluate performance of the SFD test design and to establish operational requirements for the PBF loop

  18. Fuel leak testing performance at NPP Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Slugen, V.; Krnac, S.; Smiesko, I.

    1995-01-01

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR's. 1 tab., 2 figs., 3 refs

  19. Fuel leak testing performance at NPP Jaslovske Bohunice

    Energy Technology Data Exchange (ETDEWEB)

    Slugen, V; Krnac, S [Slovak Technical Univ., Bratislava (Slovakia); Smiesko, I [Nuclear Powr Plant EBO, Jaslovske Bohuce (Slovakia)

    1996-12-31

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR`s. 1 tab., 2 figs., 3 refs.

  20. Irradiation Testing of TRISO-Coated Particle Fuel in Korea

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

    2014-01-01

    In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

  1. Thermohydraulic tests in nuclear fuel model

    International Nuclear Information System (INIS)

    Ladeira, L.C.D.; Navarro, M.A.

    1984-01-01

    The main experimental works performed in the Thermohydraulics Laboratory of the NUCLEBRAS Nuclear Technology Development Center, in the field of thermofluodynamics are briefly described. These works include the performing of steady-state flow tests in single tube test sections, and the design and construction of a rod bundle test section, which will be also used for those kind of testes. Mention is made of the works to be performed in the near future, related to steady-state and transient flow tests. (Author) [pt

  2. Power Burst Facility severe-fuel-damage test program

    International Nuclear Information System (INIS)

    McCardell, R.K.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

  3. Distilling hydrocarbon oils

    Energy Technology Data Exchange (ETDEWEB)

    Schulze, J E

    1923-03-19

    In distilling mineral oils such as petroleum, shale oil, distillates and topped or residual oils, particularly to obtain lubricating oils, the distillation is carried out under reduced pressures below an absolute pressure of 25 mm. of mercury and preferably below about 5 mm. of mercury, and the distillate is collected in fractions determined by the physical characteristics, such as viscosity, flash point, fire point, etc. Superheated steam may be passed through the liquid during distillation. A horizontal cylindrical still provided with cross braces and peripheral ribs interrupted at the base is connected through a condensing coil immersed in a steam chest and a baffled chamber with distillate receiver and is evacuated by a pump. Steam from a boiler and superheater is injected into the still through a perforated pipe. Steam and light oil vapors passing from the chamber are condensed in a coil.

  4. Multipartite nonlocality distillation

    International Nuclear Information System (INIS)

    Hsu, Li-Yi; Wu, Keng-Shuo

    2010-01-01

    The stronger nonlocality than that allowed in quantum theory can provide an advantage in information processing and computation. Since quantum entanglement is distillable, can nonlocality be distilled in the nonsignalling condition? The answer is positive in the bipartite case. In this article the distillability of the multipartite nonlocality is investigated. We propose a distillation protocol solely exploiting xor operations on output bits. The probability-distribution vectors and matrix are introduced to tackle the correlators. It is shown that only the correlators with extreme values can survive the distillation process. As the main result, the amplified nonlocality cannot maximally violate any Bell-type inequality. Accordingly, a distillability criterion in the postquantum region is proposed.

  5. Design verification testing for fuel element type CAREM

    International Nuclear Information System (INIS)

    Martin Ghiselli, A.; Bonifacio Pulido, K.; Villabrille, G.; Rozembaum, I.

    2013-01-01

    The hydraulic and hydrodynamic characterization tests are part of the design verification process of a nuclear fuel element prototype and its components. These tests are performed in a low pressure and temperature facility. The tests requires the definition of the simulation parameters for setting the test conditions, the results evaluation to feedback mathematical models, extrapolated the results to reactor conditions and finally to decide the acceptability of the tested prototype. (author)

  6. New JMTR irradiation test plan on fuels and materials

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

    2009-01-01

    In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

  7. Fouling in Membrane Distillation, Osmotic Distillation and Osmotic Membrane Distillation

    Directory of Open Access Journals (Sweden)

    Mourad Laqbaqbi

    2017-03-01

    Full Text Available Various membrane separation processes are being used for seawater desalination and treatment of wastewaters in order to deal with the worldwide water shortage problem. Different types of membranes of distinct morphologies, structures and physico-chemical characteristics are employed. Among the considered membrane technologies, membrane distillation (MD, osmotic distillation (OD and osmotic membrane distillation (OMD use porous and hydrophobic membranes for production of distilled water and/or concentration of wastewaters for recovery and recycling of valuable compounds. However, the efficiency of these technologies is hampered by fouling phenomena. This refers to the accumulation of organic/inorganic deposits including biological matter on the membrane surface and/or in the membrane pores. Fouling in MD, OD and OMD differs from that observed in electric and pressure-driven membrane processes such electrodialysis (ED, membrane capacitive deionization (MCD, reverse osmosis (RO, nanofiltration (NF, ultrafiltration (UF, microfiltration (MF, etc. Other than pore blockage, fouling in MD, OD and OMD increases the risk of membrane pores wetting and reduces therefore the quantity and quality of the produced water or the concentration efficiency of the process. This review deals with the observed fouling phenomena in MD, OD and OMD. It highlights different detected fouling types (organic fouling, inorganic fouling and biofouling, fouling characterization techniques as well as various methods of fouling reduction including pretreatment, membrane modification, membrane cleaning and antiscalants application.

  8. Distillation of bituminous shale

    Energy Technology Data Exchange (ETDEWEB)

    Seguin, M

    1875-02-16

    The retort with its accessories constitutes a distillation apparatus for shale composed of a cylindrical, vertical, fixed, tubular, and of ring form metal retort. Also it is comprised of a special hearth of large dimensions in the form of a circular pocket receiving from the retort as heating agent the distilled shale and emitting by radiation the heat that makes the distillation apparatus for the shale act.

  9. Stand for visual ultrasonic testing of spent fuel

    International Nuclear Information System (INIS)

    Czajkowski, W.; Borek-Kruszewska, E.

    2001-01-01

    A stand for visual and ultrasonic testing of spent fuel, constructed under Strategic Governmental Programme for management of spent fuel and radioactive waste, is presented in the paper. The stand, named 'STEND-1', built up at the Institute of Atomic Energy in Swjerk, is appointed for underwater visual testing of spent fuel elements type MR6 and WWR by means of TV-CCD camera and image processing system and for ultrasonic scanning of external surface of these elements by means of video scan immersion transducer and straight UHT connector. 'STEND-1' is built using flexible in use, high-tensile, anodized aluminum profiles. All the profiles feature longitudinal grooves to accommodate connecting elements and for the attachment of accessories at any position. They are also characterised by straight-through core bores for use with standard fastening elements and to accommodate accessory components. Stand, equipped with automatic control and processing system based on personal computer, may be manually or automatically controlled. Control system of movements of the camera in the vertical axis and rotational movement of spent fuel element permits to fix chosen location of fuel element with accuracy better than 0.1 mm. High resolution of ultrasonic method allows to record damages of outer surface of order 0.1 mm. The results of visual testing of spent fuel are recorded on video tape and then may be stored on the hard disc of the personal computer and presented in shape of photo or picture. Only selected damage surfaces of spent fuel elements are tested by means of ultrasonic scanning. All possibilities of the stand and results of visual testing of spent fuel type WWR are presented in the paper. (author)

  10. Distillation modeling for a uranium refining process

    Energy Technology Data Exchange (ETDEWEB)

    Westphal, B.R.

    1996-03-01

    As part of the spent fuel treatment program at Argonne National Laboratory, a vacuum distillation process is being employed for the recovery of uranium following an electrorefining process. Distillation of a salt electrolyte, containing a eutectic mixture of lithium and potassium chlorides, from uranium is achieved by a simple batch operation and is termed {open_quotes}cathode processing{close_quotes}. The incremental distillation of electrolyte salt will be modeled by an equilibrium expression and on a molecular basis since the operation is conducted under moderate vacuum conditions. As processing continues, the two models will be compared and analyzed for correlation with actual operating results. Possible factors that may contribute to aberrations from the models include impurities at the vapor-liquid boundary, distillate reflux, anomalous pressure gradients, and mass transport phenomena at the evaporating surface. Ultimately, the purpose of either process model is to enable the parametric optimization of the process.

  11. Distillation modeling for a uranium refining process

    International Nuclear Information System (INIS)

    Westphal, B.R.

    1996-01-01

    As part of the spent fuel treatment program at Argonne National Laboratory, a vacuum distillation process is being employed for the recovery of uranium following an electrorefining process. Distillation of a salt electrolyte, containing a eutectic mixture of lithium and potassium chlorides, from uranium is achieved by a simple batch operation and is termed open-quotes cathode processingclose quotes. The incremental distillation of electrolyte salt will be modeled by an equilibrium expression and on a molecular basis since the operation is conducted under moderate vacuum conditions. As processing continues, the two models will be compared and analyzed for correlation with actual operating results. Possible factors that may contribute to aberrations from the models include impurities at the vapor-liquid boundary, distillate reflux, anomalous pressure gradients, and mass transport phenomena at the evaporating surface. Ultimately, the purpose of either process model is to enable the parametric optimization of the process

  12. Irradiation testing of coated particle fuel at Hanaro

    International Nuclear Information System (INIS)

    Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

    2014-01-01

    TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

  13. Fuel integrity project: analysis of light water reactor fuel rods test results

    Energy Technology Data Exchange (ETDEWEB)

    Dallongeville, M.; Werle, J. [COGEMA Logistics (AREVA Group) (France); McCreesh, G. [BNFL Nuclear Sciences and Technology Services (United Kingdom)

    2004-07-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  14. Fuel integrity project: analysis of light water reactor fuel rods test results

    International Nuclear Information System (INIS)

    Dallongeville, M.; Werle, J.; McCreesh, G.

    2004-01-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  15. Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

  16. Improvement of test methodology for evaluating diesel fuel stability

    Energy Technology Data Exchange (ETDEWEB)

    Gutman, M.; Tartakovsky, L.; Kirzhner, Y.; Zvirin, Y. [Internal Combustion Engines Lab., Haifa (Israel); Luria, D. [Fuel Authority, Tel Aviv (Israel); Weiss, A.; Shuftan, M. [Israel Defence Forces, Tel Aviv (Israel)

    1995-05-01

    The storage stability of diesel fuel has been extensively investigated for many years under laboratory conditions. Although continuous efforts have been made to improve testing techniques, there does not yet exist a generally accepted correlation between laboratory methods (such as chemical analysis of the fuel) and actual diesel engine tests. A testing method was developed by the Technion Internal Combustion Engines Laboratory (TICEL), in order to address this problem. The test procedure was designed to simulate diesel engine operation under field conditions. It is based on running a laboratory-modified single cylinder diesel engine for 50 h under cycling operating conditions. The overall rating of each test is based on individual evaluation of the deposits and residue formation in the fuel filter, nozzle body and needle, piston head, piston rings, exhaust valve, and combustion chamber (six parameters). Two methods for analyzing the test results were used: objective, based on measured data, and subjective, based on visual evaluation results of these deposits by a group of experts. Only the residual level in the fuel filter was evaluated quantitatively by measured results. In order to achieve higher accuracy of the method, the test procedure was improved by introducing the measured results of nozzle fouling as an additional objective evaluating (seventh) parameter. This factor is evaluated on the basis of the change in the air flow rate through the nozzle before and after the complete engine test. Other improvements in the method include the use of the nozzle assembly photograph in the test evaluation, and representation of all seven parameters on a continuous scale instead of the discrete scale used anteriorly, in order to achieve higher accuracy. This paper also contains the results obtained by application of this improved fuel stability test for a diesel fuel stored for a five-year period.

  17. Updated FY12 Ceramic Fuels Irradiation Test Plan

    International Nuclear Information System (INIS)

    Nelson, Andrew T.

    2012-01-01

    The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

  18. The Use of Fuel Gas as Stripping Medium in Atmospheric Distillation of Crude Oil L’utilisation de gaz combustible comme moyen d’extraction des fractions légères en distillation atmosphérique du pétrole brut

    Directory of Open Access Journals (Sweden)

    Plellis-Tsaltakis C.

    2011-09-01

    Full Text Available Stripping of petroleum fractions aims to remove the light ends that spoil some of their properties, such as the flash point. Stripping usually employs steam for that purpose. Except for steam, other substances can perform the same function, among them light hydrocarbons. In this article, we investigate the use of refinery fuel gas as stripping medium for a crude oil atmospheric distillation unit. L’extraction des fractions légères du pétrole vise à enlever les produits nuisibles à certaines propriétés comme le point d’éclair. Cette opération utilise habituellement de la vapeur d’eau. D’autres substances peuvent remplir la même fonction, comme les hydrocarbures légers. Dans cet article, on étudie l’utilisation du gaz combustible de la raffinerie comme moyen d’extraction des fractions légères dans une unité de distillation atmosphérique.

  19. Regenerative adsorption distillation system

    KAUST Repository

    Ng, Kim Choon

    2013-12-26

    There is provided a regenerative adsorption distillation system comprising a train of distillation effects in fluid communication with each other. The train of distillation effects comprises at least one intermediate effect between the first and last distillation effects of the train, each effect comprising a vessel and a condensing tube for flow of a fluid therein. The system further comprises a pair of adsorption-desorption beds in vapour communication with the last effect and at least one intermediate effect, wherein the beds contain an adsorbent that adsorbs vapour from the last effect and transmits desorbed vapour into at least one of the intermediate effect.

  20. Regenerative adsorption distillation system

    KAUST Repository

    Ng, Kim Choon; Thu, Kyaw; Amy, Gary; Chunggaze, Mohammed; Al-Ghasham, Tawfiq

    2013-01-01

    There is provided a regenerative adsorption distillation system comprising a train of distillation effects in fluid communication with each other. The train of distillation effects comprises at least one intermediate effect between the first and last distillation effects of the train, each effect comprising a vessel and a condensing tube for flow of a fluid therein. The system further comprises a pair of adsorption-desorption beds in vapour communication with the last effect and at least one intermediate effect, wherein the beds contain an adsorbent that adsorbs vapour from the last effect and transmits desorbed vapour into at least one of the intermediate effect.

  1. BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

    International Nuclear Information System (INIS)

    Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

    2002-01-01

    Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

  2. Characterization and testing of monolithic RERTR fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.; Jue, J.F.; Burkes, D.E. [Idaho National Lab., Idaho Falls, ID (United States)

    2007-07-01

    Monolithic fuel plates are being developed as a LEU (low enrichment uranium) fuel for application in research reactors throughout the world. These fuel plates are comprised of a U-Mo alloy foil encased in aluminum alloy cladding. Three different fabrication techniques have been looked at for producing monolithic fuel plates: hot isostatic pressing (HIP), transient liquid phase bonding (TLPB), and friction stir welding (FSW). Of these three techniques, HIP and FSW are currently being emphasized. As part of the development of these fabrication techniques, fuel plates are characterized and tested to determine properties like hardness and the bond strength at the interface between the fuel and cladding. Testing of HIP-made samples indicates that the foil/cladding interaction behavior depends on the Mo content in the UMo foil, the measured hardness values are quite different for the fuel, cladding, and interaction zone phase and Ti, Zr and Nb are the most effective diffusion barriers. For FSW samples, there is a dependence of the bond strength at the foil/cladding interface on the type of tool that is employed for performing the actual FSW process. (authors)

  3. Evaluation of the linear power of HANARO test fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Sung; Seo, C. G.; Lee, B. C.; Kim, H. R

    2001-02-01

    The HANARO fuel was developed by AECL and it is configured in a bundle of rods containing uranium silicide. AECL has conducted a variety of tests using specimen in order to achieve its qualification and licensing and the highest linear power was evaluated to be 112.8kW/m. In design stage of HANARO, the best estimated maximum linear power at hot spot was found to occur in the transition core from the initial to the equilibrium and its value was 108kW/m, which exceeds 112.8kW/m if the physics uncertainty of the HANARO nuclear design model is taken into account. Consequently, the licensing body issued the conditional permit to operate HANARO and the fuel integrity at the linear power higher than 112.8kW/m was requested to be confirmed through irradiation tests by realizing its repeatability. Hereby, KAERI designed uninstrumented and instrumented test fuel bundles and conducted their burnup tests. In parallel with the tests, the nuclear design model has been revised and updated to enable us to pursue the pin-by-pin power history. This report describes the best estimated power history of the test fuel bundles using the revised model. In conclusion, HANARO fuel keeps its integrity at power condition greater than 120kW/m.

  4. HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    1998-03-01

    As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

  5. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    International Nuclear Information System (INIS)

    Katayama, Y.B.

    1979-04-01

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

  6. Design of experiments for test of fuel element reliability

    International Nuclear Information System (INIS)

    Boehmert, J.; Juettner, C.; Linek, J.

    1989-01-01

    Changes of fuel element design and modifications of the operational conditions have to be tested in experiments and pilot projects for nuclear safety. Experimental design is an useful statistical method minimizing costs and risks for this procedure. The main problem of our work was to investigate the connection between failure rate of fuel elements, sample size, confidence interval, and error probability. Using the statistic model of the binomial distribution appropriate relations were derived and discussed. A stepwise procedure based on a modified sequential analysis according to Wald was developed as a strategy of introduction for modifications of the fuel element design and of the operational conditions. (author)

  7. Sipping test on a failed MTR fuel element

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

    2002-01-01

    This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

  8. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

  9. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    Kohn, E.; Wright, M.G.

    1976-11-01

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  10. Space reactor fuel element testing in upgraded TREAT

    International Nuclear Information System (INIS)

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

  11. Testing of a transport cask for research reactor spent fuel

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2011-01-01

    Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

  12. Space reactor fuel element testing in upgraded TREAT

    Science.gov (United States)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

  13. Modeling and Simulation of a Nuclear Fuel Element Test Section

    Science.gov (United States)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  14. Performance and emission study on waste cooking oil biodiesel and distillate blends for microturbine application

    Directory of Open Access Journals (Sweden)

    Ee Sann Tan

    2015-11-01

    Full Text Available Biodiesel is defined as domestic renewable energy resource, which can be derived from natural oils through the transesterification. The implementation of biodiesel is essential due to the energy depletion crisis and the impact on exacerbating environment caused by rapid consumption of conventional diesel. Waste cooking oil (WCO was used as the raw material to produce biodiesel in order to reduce wastes polluting the environment. This paper studies the technical potential of WCO biodiesel to be used as an alternative fuel for microturbine. The ASTM D6751 and ASTM D2881 standards were selected as references to evaluate the compatibility with distillate to be used as a microturbine fuel. The performance and emission tests were conducted employing a 30 kW microturbine, without any modification, using biodiesel and distillate blends up to maximum of 20% biodiesel mixing ratio. It was found that the thermal efficiency peaked at 20% biodiesel blend with distillate, despite the fact that biodiesel had a lower calorific value and a higher fuel consumption. The emission test results showed reduction of CO emission by increasing the WCO biodiesel mixing ratio, while NOx emission was dependent on the exhaust gas temperature. In conclusion, biodiesel derived from WCO has the potential to substitute distillate in the microturbine application.

  15. 16 x 16 Vantage+ Fuel Assembly Flow Vibrational Testing

    International Nuclear Information System (INIS)

    Chambers, Martin; Kurincic, Bojan

    2014-01-01

    Nuklearna Elektrarna Krsko (NEK) has experienced leaking fuel after increasing the cycle duration to 18 months. The leaking fuel mechanism has predominantly been consistent over multiple cycles and is typically observed in highly irradiated Fuel Assemblies (FA) after around 4 years of continuous operation that were located at the core periphery (baffle). The cause of the leaking fuel is due to Grid-To-Rod-Fretting (GRTF) and occasional debris fretting. NEK utilises a 16x16 Vantage+ FA design with all Inconel structural mixing vane grids (8 in total), Zirlo thimbles, Integral Fuel Burnable Absorber (IFBA) rods with enriched ZrB2, enriched Annular Blanket, Debris Filter Bottom Nozzle (DFBN), Removable Top Nozzle (RTN) and Zirlo fuel cladding material with a high burnup capability of 60 GWD/MTU. Numerous design and operational changes are thought to have reduced the original 16x16 FA design margin to fretting resistance of either vibration or its wear work rate, such as significant power uprate (spring force loss, rod creep down...), operational cycle duration increase from 12 to 18 months (increasing residence time as well as lead FA and fuel rod burnup values), Reactor Coolant System flow increase (increased vibration), removal of Thimble Plugs (increased bypass flow, increased vibration) and Zirc-4 to Zirlo cladding change (decreasing wear work rate). The fuel rod to grid spring as well as dimple contact areas are relatively smaller than other FA designs that exhibit good in-reactor fretting performance. A FA design change project to address the small rod to dimple / spring contact area and utilise fuel cladding oxide coating is currently being pursued with the fuel supplier. The FA vibrational properties are very important to the in-reactor FA performance and reliability. The 16x16 Vantage+ vibrational testing was performed with a full size FA in the Fuel Assembly Compatibility Testing (FACTS) loop that is able to provide full flow rates at elevated temperature

  16. Fundamental study on the salt distillation from the mixtures of rare earth precipitates and LiCl-KCl eutectic salt

    International Nuclear Information System (INIS)

    Yang, H. C.; Eun, H. C.; Cho, Y. Z.; Lee, H. S.; Kim, I. T.

    2008-01-01

    An electrorefining process of spent nuclear fuel generates waste salt containing some radioactive metal chlorides. The most effective method to reduce salt waste volume is to separate radioactive metals from non-radioactive salts. A promising approach is to change radioactive metal chlorides into salt-insoluble oxides by an oxygen sparging. Following this, salt distillation process is available to effectively separate the precipitated particulate metal oxides from salt. This study investigated the distillation rates of LiCl-KCl eutectic salt under different vacuums at elevated temperatures. The first part study investigated distillation rates of eutectic salt under different vacuums at high temperatures by using thermo-gravimetric furnace system. In the second part, we tested the removal of eutectic salt from the RE precipitates by using the laboratory vacuum distillation furnace system. Investigated variables were the temperature of mixture, the degree of vacuum and the time

  17. Welding of metallic fuel elements for the irradiation test in JOYO. Preliminary tests and welding execution tests (Joint research)

    International Nuclear Information System (INIS)

    Kikuchi, Hironobu; Nakamura, Kinya; Iwai, Takashi; Arai, Yasuo

    2009-10-01

    Irradiation tests of metallic fuels elements in fast test reactor JOYO are planned under the joint research of Japan Atomic Energy Agency (JAEA) and Central Research Institute of Electric Power Industry (CRIEPI). Six U-Pu-Zr fuel elements clad with ferritic martensitic steel are fabricated in Plutonium Fuel Research Facility (PFRF) of JAEA-Oarai for the first time in Japan. In PFRF, the procedures of fabrication of the fuel elements were determined and the test runs of the equipments were carried out before the welding execution tests for the fuel elements. Test samples for confirming the welding condition between the cladding tube and top and bottom endplugs were prepared, and various test runs were carried out before the welding execution tests. As a result, the welding conditions were finalized by passing the welding execution tests. (author)

  18. Test plan for Series 3 NNWSI spent fuel leaching/dissolution tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1986-04-01

    The Series 3 tests will differ from the Series 2 tests in that the Series 3 tests will be run at 85 0 C (J-13 water) in sealed 304 stainless steel (SS) test vessels. The current NNWSI reference spent fuel container material is 304L SS. The candidate NNWSI repository horizon is above the water table, and 95 0 C (boiling temperature at the repository elevation) is the maximum liquid water temperature expected to contact spent fuel in the repository

  19. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  20. Distillation plant for tritium enrichment in metallic lithium

    International Nuclear Information System (INIS)

    Barnert, E.; Butzek, D.; Cordewiner, J.; Heinrichs, E.

    1984-06-01

    To close the external fuel cycle of fusion reactors, the tritium obtained from lithium must be separated off. One way of doing this is by high-temperature distillation and subsequent permeation. The construction of high-temperature distillation plant is described. For the time being, deuterium is processed instead of tritium. (orig.) [de

  1. Vacuum distillation of plutonium pyrochemical salts

    International Nuclear Information System (INIS)

    Bourges, Gilles; Faure, S.; Fiers, B.; Saintignon, S.; Lemoine, O.; Cardona-Barrau, D.; Devillard, D.

    2012-01-01

    A pyrochemical process is developed to upgrade the safety of plutonium spent salts interim storage. The feed material, consisting of alkali or alkali-earth chlorides containing various Pu and Am species, is first oxidized to convert the actinides into oxides. Then the chlorides are removed by vacuum distillation which requires temperature from 750 degrees C to 1100 degrees C. After a comprehensive R and D program, full-scale equipment was built to test the distillation of active salts. Tests with NaCl/KCl oxidized spent salt give decontamination factor of chlorides higher than 20000. The distilled salt meets the radiologic requirements to be discarded as low level waste. (authors)

  2. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  3. Municipal solid waste combustion: Fuel testing and characterization

    Energy Technology Data Exchange (ETDEWEB)

    Bushnell, D.J.; Canova, J.H.; Dadkhah-Nikoo, A.

    1990-10-01

    The objective of this study is to screen and characterize potential biomass fuels from waste streams. This will be accomplished by determining the types of pollutants produced while burning selected municipal waste, i.e., commercial mixed waste paper residential (curbside) mixed waste paper, and refuse derived fuel. These materials will be fired alone and in combination with wood, equal parts by weight. The data from these experiments could be utilized to size pollution control equipment required to meet emission standards. This document provides detailed descriptions of the testing methods and evaluation procedures used in the combustion testing and characterization project. The fuel samples will be examined thoroughly from the raw form to the exhaust emissions produced during the combustion test of a densified sample.

  4. Thermal test and analysis of a spent fuel storage cask

    International Nuclear Information System (INIS)

    Yamakawa, H.; Gomi, Y.; Ozaki, S.; Kosaki, A.

    1993-01-01

    A thermal test simulated with full-scale cask model for the normal storage was performed to verify the storage skill of the spent fuels of the cask. The maximum temperature at each point in the test was lower than the allowable temperature. The integrity of the cask was maintained. It was observed that the safety of containment system was also kept according to the check of the seal before and after the thermal test. Therefore it was shown that using the present skill, it is possible to store spent fuels in the dry-type cask safely. Moreover, because of the good agreement between analysis and experimental results, it was shown that the analysis model was successfully established to estimate the temperature distribution of the fuel cladding and the seal portion. (J.P.N.)

  5. Degradation mechanisms and accelerated testing in PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Borup, Rodney L [Los Alamos National Laboratory; Mukundan, Rangachary [Los Alamos National Laboratory

    2010-01-01

    The durability of PEM fuel cells is a major barrier to the commercialization of these systems for stationary and transportation power applications. Although there has been recent progress in improving durability, further improvements are needed to meet the commercialization targets. Past improvements have largely been made possible because of the fundamental understanding of the underlying degradation mechanisms. By investigating component and cell degradation modes; defining the fundamental degradation mechanisms of components and component interactions new materials can be designed to improve durability. Various factors have been shown to affect the useful life of PEM fuel cells. Other issues arise from component optimization. Operational conditions (such as impurities in either the fuel and oxidant stream), cell environment, temperature (including subfreezing exposure), pressure, current, voltage, etc.; or transient versus continuous operation, including start-up and shutdown procedures, represent other factors that can affect cell performance and durability. The need for Accelerated Stress Tests (ASTs) can be quickly understood given the target lives for fuel cell systems: 5000 hours ({approx} 7 months) for automotive, and 40,000 hrs ({approx} 4.6 years) for stationary systems. Thus testing methods that enable more rapid screening of individual components to determine their durability characteristics, such as off-line environmental testing, are needed for evaluating new component durability in a reasonable turn-around time. This allows proposed improvements in a component to be evaluated rapidly and independently, subsequently allowing rapid advancement in PEM fuel cell durability. These tests are also crucial to developers in order to make sure that they do not sacrifice durability while making improvements in costs (e.g. lower platinum group metal [PGM] loading) and performance (e.g. thinner membrane or a GDL with better water management properties). To

  6. Catalytic distillation process

    Science.gov (United States)

    Smith, L.A. Jr.

    1982-06-22

    A method is described for conducting chemical reactions and fractionation of the reaction mixture comprising feeding reactants to a distillation column reactor into a feed zone and concurrently contacting the reactants with a fixed bed catalytic packing to concurrently carry out the reaction and fractionate the reaction mixture. For example, a method for preparing methyl tertiary butyl ether in high purity from a mixed feed stream of isobutene and normal butene comprising feeding the mixed feed stream to a distillation column reactor into a feed zone at the lower end of a distillation reaction zone, and methanol into the upper end of said distillation reaction zone, which is packed with a properly supported cationic ion exchange resin, contacting the C[sub 4] feed and methanol with the catalytic distillation packing to react methanol and isobutene, and concurrently fractionating the ether from the column below the catalytic zone and removing normal butene overhead above the catalytic zone.

  7. Irradiation test plan of the simulated DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M. S.; Kim, B. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    Simulated DUPIC fuel had been irradiated from Aug. 4, 1999 to Oct. 4 1999, in order to produce the data of its in-core behavior, to verify the design of DUPIC non-instrumented capsule developed, and to ensure the irradiation requirements of DUPIC fuel at HANARO. The welding process was certified for manufacturing the mini-element, and simulated DUPIC fuel rods were manufactured with simulated DUPIC pellets through examination and test. The non-instrumented capsule for a irradiation test of DUPIC fuel has been designed and manufactured referring to the design specification of the HANARO fuel. This is to be the design basis of the instrumented capsule under consideration. The verification experiment, whether the capsule loaded in the OR4 hole meet the HANARO requirements under the normal operation condition, as well as the structural analysis was carried out. The items for this experiment were the pressure drop test, vibration test, integrity test, et. al. It was noted that each experimental result meet the HANARO operational requirements. For the safety analysis of the DUPIC non-instrumented capsule loaded in the HANARO core, the nuclear/mechanical compatibility, thermodynamic compatibility, integrity analysis of the irradiation samples according to the reactor condition as well as the safety analysis of the HANARO were performed. Besides, the core reactivity effects were discussed during the irradiation test of the DUPIC capsule. The average power of each fuel rod in the DUPIC capsule was calculated, and maximal linear power reflecting the axial peaking power factor from the MCNP results was evaluated. From these calculation results, the HANARO core safety was evaluated. At the end of this report, similar overseas cases were introduced. 9 refs., 16 figs., 10 tabs. (Author)

  8. Analytic tests and their relation to jet fuel thermal stability

    Energy Technology Data Exchange (ETDEWEB)

    Heneghan, S.P.; Kauffman, R.E. [Univ. of Dayton Research Institute, OH (United States)

    1995-05-01

    The evaluation of jet fuel thermal stability (TS) by simple analytic procedures has long been a goal of fuels chemists. The reason is obvious: if the analytic chemist can determine which types of material cause his test to respond, the refiners will know which materials to remove to improve stability. Complicating this quest is the lack of an acceptable quantitative TS test with which to compare any analytic procedures. To circumvent this problem, we recently compiled the results of TS tests for 12 fuels using six separate test procedures. The results covering a range of flow and temperature conditions show that TS is not as dependent on test conditions as previously thought. Also, comparing the results from these tests with several analytic procedures shows that either a measure of the number of phenols or the total sulfur present in jet fuels is strongly indicative of the TS. The phenols have been measured using a cyclic voltammetry technique and the polar material by gas chromatography (atomic emission detection) following a solid phase extraction on silica gel. The polar material has been identified as mainly phenols (by mass spectrometry identification). Measures of the total acid number or peroxide concentration have little correlation with TS.

  9. Status on the construction of the fuel irradiation test facility

    International Nuclear Information System (INIS)

    Park, Kook Nam; Sim, Bong Shick; Lee, Chung Young; Yoo, Seong Yeon

    2005-01-01

    As a facility to examine general performance of nuclear fuel under irradiation condition in HANARO, Fuel Test Loop(FTL) has been developed which can accommodate 3 fuel pins at the core irradiation hole(IR1 hole) taking consideration user's test requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of- Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. 3-Pin FTL Conceptual design was set up in 2001 and had completed detail design including a design requirement and basic Piping and Instrument Diagram (P and ID) in 2004. The safety analysis report was prepared and submitted in early 2005 to the regulatory body(KINS) for review and approval of FTL. In 2005, the development team is going to purchase and manufacture hardware and make a contract for construction work. In 2006, the development team is going to install an FTL system performance test shall be done as a part of commissioning. After a 3-Pin FTL development which is expected to be finished by the 2007, FTL will be used for the irradiation test of the new PWR-type fuel and the usage of HANARO will be enhanced

  10. Crash testing of nuclear fuel shipping containers

    International Nuclear Information System (INIS)

    Jefferson, R.M.; Yoshimura, H.R.

    1977-08-01

    In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. The program encompasses the following distinct major efforts. The first of these utilizes computational methods for predicting the effects of the accident environment and, subsequently, to calculate the damage incurred by a container as the result of such an accident. The second phase involves the testing of 1 / 8 -scale models of transportation systems. Through the use of instrumentation and high-speed motion photography the accident environments and physical damage mechanisms are studied in detail. After correlating the results of these first two phases, a full scale event involving representative hardware is conducted. To date two of the three selected test scenarios have been completed. Results of the program to this point indicate that both computational techniques and scale modeling are viable engineering approaches to studying accident environments and physical damage to shipping casks

  11. Crash testing of nuclear fuel shipping containers

    International Nuclear Information System (INIS)

    Jefferson, R.M.; Yoshimura, H.R.

    1977-12-01

    In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. This program, which began in 1975, encompasses the following distinct major efforts. The first of these utilizes computational methods for predicting the effects of the accident environment and, subsequently, to calculate the damage incurred by a container as the result of such an accident. The second phase involves the testing of 1 / 8 -scale models of transportation systems. Through the use of instrumentation and high-speed motion photography, the accident environments and physical damage mechanisms are studied in detail. After correlating the results of these first two phases, a full scale event involving representative hardware is conducted. To date two of the three selected test scenarios have been completed. Results of the program to this point indicate that both computational techniques and scale modeling are viable engineering approaches to studying accident environments and physical damage to shipping casks

  12. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  13. Real life testing of a Hybrid PEM Fuel Cell Bus

    Science.gov (United States)

    Folkesson, Anders; Andersson, Christian; Alvfors, Per; Alaküla, Mats; Overgaard, Lars

    Fuel cells produce low quantities of local emissions, if any, and are therefore one of the most promising alternatives to internal combustion engines as the main power source in future vehicles. It is likely that urban buses will be among the first commercial applications for fuel cells in vehicles. This is due to the fact that urban buses are highly visible for the public, they contribute significantly to air pollution in urban areas, they have small limitations in weight and volume and fuelling is handled via a centralised infrastructure. Results and experiences from real life measurements of energy flows in a Scania Hybrid PEM Fuel Cell Concept Bus are presented in this paper. The tests consist of measurements during several standard duty cycles. The efficiency of the fuel cell system and of the complete vehicle are presented and discussed. The net efficiency of the fuel cell system was approximately 40% and the fuel consumption of the concept bus is between 42 and 48% lower compared to a standard Scania bus. Energy recovery by regenerative braking saves up 28% energy. Bus subsystems such as the pneumatic system for door opening, suspension and brakes, the hydraulic power steering, the 24 V grid, the water pump and the cooling fans consume approximately 7% of the energy in the fuel input or 17% of the net power output from the fuel cell system. The bus was built by a number of companies in a project partly financed by the European Commission's Joule programme. The comprehensive testing is partly financed by the Swedish programme "Den Gröna Bilen" (The Green Car). A 50 kW el fuel cell system is the power source and a high voltage battery pack works as an energy buffer and power booster. The fuel, compressed hydrogen, is stored in two high-pressure stainless steel vessels mounted on the roof of the bus. The bus has a series hybrid electric driveline with wheel hub motors with a maximum power of 100 kW. Hybrid Fuel Cell Buses have a big potential, but there are

  14. Development of a non-engine fuel injector deposit test for alternative fuels (ENIAK-project)

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Hajo; Pohland vom Schloss, Heide [OWI - Oel Waerme Institut GmbH, Herzogenrath (Germany)

    2013-06-01

    Deposit formation in and on the injectors of diesel engines may lead to injector malfunction, resulting in a loss in power, rough engine operation and poor emission levels. Poor Biodiesel quality, contamination with copper and zinc as well as undesired reactions between (several) additives and biodiesel components are known causes for nozzle fouling. Therefore, good housekeeping when using biodiesel is required, and all additives have to pass a no-harm test concerning injector fouling. The standard fouling tests are two engine tests: The XUD9-test (CEC F-23-01) and the DW-10-test (CEC DF 98-08). The XUD9 is a cost efficient, fast and proven testing method. It uses, however, an obsolete indirect injection diesel engine and cannot reproduce internal diesel injector deposits (IDID). The newer DW10 test is complex, costly and designed for high stress. This reduces the engine life and leads to a fuel consumption of approximately 1,000 1 per test, both contributing to the high costs of the test. The ENIAK-Project is funded by the FNR (''Fachagentur Nachwachsende Rohstoffe'', Agency for Renewable Resources) and conducted in cooperation with AGQM, ASG and ERC. Its main goal is the development, assembly, commissioning, and evaluation of a non-engine fuel injector test. It uses a complete common rail system. The injection takes place in a self-designed reactor instead of an engine, and the fuel is not combusted, but re-condensed and pumped in a circle, leading to a low amount of fuel required. If the test method proves to be as reliable as expected, it can be used as an alternative test method for injector fouling with low requirements regarding infrastructure on the testing site and sample volume. (orig.)

  15. Spent fuel test. Climax data acquisition system integration report

    International Nuclear Information System (INIS)

    Nyholm, R.A.; Brough, W.G.; Rector, N.L.

    1982-06-01

    The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures

  16. Spent Fuel Test - Climax data acquisition system operations manual

    International Nuclear Information System (INIS)

    Nyholm, R.A.

    1983-01-01

    The Spent Fuel Test-Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granite rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. The multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system (DAS) collects and processes data from more than 900 analog instruments. This report documents the software element of the LLNL developed SFT-C Data Acquisition System. It defines the operating system and hardware interface configurations, the special applications software and data structures, and support software

  17. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  18. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  19. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  20. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    Science.gov (United States)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  1. Low temperature destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    1938-07-05

    A process is given and apparatus is described for the destructive distillation at low temperature of coal, oil shale, and the like by subjection to the action of a stream of hot gases or superhearted steam, flowing in a closed circuit. Subsequent treatment of the distillation residues with a gas stream containing oxygen results in combustion of the carbon-containing material therein brings to a high temperature the solid residue, in which the process comprises subsequently contacting the hot solid residue with the fluid stream effecting the distillation.

  2. Fuel staging tests at the Kymijaervi power plant

    International Nuclear Information System (INIS)

    Kivelae, M.; Rotter, H.; Virkki, J.

    1990-01-01

    The aim of this study was to measure nitrogen oxide (NO x ) emissions and find the methods to reduce them in plants using coal and natural gas as fuel. The tests involved were made at the Kymijaervi Power Plant, Lahti, Finland. Coal and natural gas was used alone or mixed. With natural gas when using flue gas recirculation, the NO x emission level dropped from 330 mg/m 3 down to 60 mg/m 3 . A negative side effect was that the flue gas temperature increased. At coal combustion and staged combustion, the flue gas recirculation had no significant effect on the NO x emission level. At coal combustion, the staging of combustion air halved the NO x emission but the combustibles increased strongly. With fuel staging, using coal as main fuel and gas as staging fuel, the NO x emission level was decreased from 340 mg/m 3 to 170 mg/m 3 . At the same time the combustibles increased 2 %- units. Also the flue gas temperature increased a little. At the tests, the proportion of natural gas was rather high, one third of the fuel energy input, but it could not be decreased, because the gas flow ratio was already too low to ensure good mixing

  3. Test results for fuel cell operation on anaerobic digester gas

    Science.gov (United States)

    Spiegel, R. J.; Preston, J. L.

    EPA, in conjunction with ONSI, embarked on a project to define, design, test, and assess a fuel cell energy recovery system for application at anaerobic digester waste water (sewage) treatment plants. Anaerobic digester gas (ADG) is produced at these plants during the process of treating sewage anaerobically to reduce solids. ADG is primarily comprised of methane (57-66%), carbon dioxide (33-39%), nitrogen (1-10%), and a small amount of oxygen (sulfur-bearing compounds (principally hydrogen sulfide) and halogen compounds (chlorides). The project has addressed two major issues: development of a cleanup system to remove fuel cell contaminants from the gas and testing/assessing of a modified ONSI PC25 C fuel cell power plant operating on the cleaned, but dilute, ADG. Results to date demonstrate that the ADG fuel cell power plant can, depending on the energy content of the gas, produce electrical output levels close to full power (200 kW) with measured air emissions comparable to those obtained by a natural gas fuel cell. The cleanup system results show that the hydrogen sulfide levels are reduced to below 10 ppbv and halides to approximately 30 ppbv.

  4. Distilling carbonaceous material

    Energy Technology Data Exchange (ETDEWEB)

    Karrick, L C

    1926-11-02

    Coal, shale and the like are distilled in a current of superheated steam which is passed into a retort at about midway between its ends a further supply being if necessary introduced at the bottom to generate water-gas, and the coke being dry quenched in a hopper below the retort. Combustion products may also be introduced and the temperature may be varied from 950 to 1725/sup 0/F, oil, gas, resin and a residual coke having good adsorbent value, being obtained. The charge from hoppers and auxiliary hoppers is fed to retorts situated between gas and steam preheaters, the rate of downward movement being controlled by discharge rollers having arms, counterweighted rocking arms allowing the residue to be fed downwards into hoppers. Steam from a pipe is superheated in horizontal passages, and admitted through ports into the retort. Preheated fuel gas is burnt in combustion flues and passes down through vertical flues, across horizontal flues and up flues adjacent the retorts, from which by ports and flues it passes down a chamber having an air or gas preheater each having two independent systems, one discharging into the combustion chamber beneath it and the other into an adjacent chamber. Air or gas enters by pipes and after being heated in pipes is fed by ports to the chamber. The volatiles pass off through outlets leading to a main air cooled condenser and a water-cooled condenser delivering to a separating tank connecting with pipes for quenching the residues and with pipes to preheat the charge in the hoppers and maintain a gas barrier the rein. Superheated steam may also be admitted through ports to generate water-gas and increase the total volume of gases and combustion products may be introduced through ports. The upper part of the retort is made of cast iron, the high temperature parts of silica or carborundum brick, and the lower part of chrome iron or other metal.

  5. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    Delpeyroux, P.

    1964-01-01

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

  6. Prototype spent-fuel canister design, analysis, and test

    International Nuclear Information System (INIS)

    Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

    1982-03-01

    Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

  7. LOCA scenario tests of irradiated fuel rod specimens

    International Nuclear Information System (INIS)

    Scott, Harold

    2004-01-01

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

  8. Machine Vision Tests for Spent Fuel Scrap Characteristics

    International Nuclear Information System (INIS)

    BERGER, W.W.

    2000-01-01

    The purpose of this work is to perform a feasibility test of a Machine Vision system for potential use at the Hanford K basins during spent nuclear fuel (SNF) operations. This report documents the testing performed to establish functionality of the system including quantitative assessment of results. Fauske and Associates, Inc., which has been intimately involved in development of the SNF safety basis, has teamed with Agris-Schoen Vision Systems, experts in robotics, tele-robotics, and Machine Vision, for this work

  9. 40 CFR 1060.521 - How do I test fuel caps for permeation emissions?

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false How do I test fuel caps for permeation... EQUIPMENT Test Procedures § 1060.521 How do I test fuel caps for permeation emissions? If you measure a fuel tank's permeation emissions with a nonpermeable covering in place of the fuel cap under § 1060.520(b)(5...

  10. Testing of reactor fuel materials using nuclear techniques

    International Nuclear Information System (INIS)

    Khouri, M.T.F.C.

    1978-01-01

    The tests presented here apply to: the quantitative determination of uranium in the core of fuel element plates by the detection of the number of neutrons produced in photo induced reactions in uranium; the determination of 235 U proportion in uranium dioxide samples, in the form of uranyl nitrate, by the technique of the detection of tracks produced by fission fragments and in pellet samples by passive gamma spectrometry and the checking of uranium homogenization distribution in fuel plates and uranium dioxide pellets. (Author) [pt

  11. Process and apparatus for destructive distillation

    Energy Technology Data Exchange (ETDEWEB)

    Dalin, D; Hedbaeck, T J

    1951-01-10

    A process of distilling wood, coal, shale, and like materials in an externally heated retort, consists of heating the retort by burning fuel in a combustion chamber completely or partly surrounding the retort and passing a heat-absorbing medium through ducts which are mounted in or adjacent the greater part of the length of the retort walls which are so arranged as to effect a greater degree of heat extraction at one part of the retort than at another part of the retort. The zones of different heat extraction being related to the heat developed in the combustion chamber maintains the most favourable distillation temperature in all parts of the retort.

  12. Distilling shale and coal

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, H; Young, G

    1923-01-09

    In a process of recovering oil from shale or coal the material is ground and may be subjected to a cleaning or concentrating process of the kind described in Specification 153,663 after which it is distilled in a furnace as described in Specification 13,625/09 the sections of the furnace forming different temperature zones, and the rate of the passage of the material is regulated so that distillation is complete with respect to the temperature of each zone, the whole distillation being accomplished in successive stages. The vapors are taken off at each zone and superheated steam may be passed into the furnace at suitable points and the distillation terminated at any stage of the process.

  13. Low temperature distillation

    Energy Technology Data Exchange (ETDEWEB)

    1937-07-21

    To distil mineral or organic material, the material is heated by means of a hot gas entering into contact with and traversing the material in a horizontal direction. The vertical retort is charged with material from the hopper and hot gases from the furnace after traversing the boiler enter the preheating zone, pass through the slots and charge therein and are educted through connection by the blower. The charge passes downwardly to the distillation zone wherein hot gases from the heater are passed via pipe and slots through the chambers and to the pipe controlled by the dampers. These gases are recycled by the fan through the heater and chamber, that portion of the gases corresponding to the amount of vapours evolved from the distillation being evacuated via the pipe to the condensing plant not shown. Steam and/or a limited quantity of water may be admitted to the distillation zone.

  14. ROTARY FUEL INJECTION PUMP WEAR TESTING USING A 30 %/ 70% ATJ/F-24 FUEL BLEND

    Science.gov (United States)

    2017-09-30

    DD-MM-YYYY) 30-09-2017 2. REPORT TYPE Interim Report 3. DATES COVERED (From - To) September 2013 – September 2017 4. TITLE AND SUBTITLE Rotary... Corrosion Inhibitor/Lubricity Improver cm...fuels, to full scale equipment and fleet testing to determine resulting component and vehicle performance. This report covers investigation into the

  15. Irradiation testing of high density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U 2 Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions

  16. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

    1996-09-01

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

  17. Process modeling and control applied to real-time monitoring of distillation processes by near-infrared spectroscopy.

    Science.gov (United States)

    de Oliveira, Rodrigo R; Pedroza, Ricardo H P; Sousa, A O; Lima, Kássio M G; de Juan, Anna

    2017-09-08

    A distillation device that acquires continuous and synchronized measurements of temperature, percentage of distilled fraction and NIR spectra has been designed for real-time monitoring of distillation processes. As a process model, synthetic commercial gasoline batches produced in Brazil, which contain mixtures of pure gasoline blended with ethanol have been analyzed. The information provided by this device, i.e., distillation curves and NIR spectra, has served as initial information for the proposal of new strategies of process modeling and multivariate statistical process control (MSPC). Process modeling based on PCA batch analysis provided global distillation trajectories, whereas multiset MCR-ALS analysis is proposed to obtain a component-wise characterization of the distillation evolution and distilled fractions. Distillation curves, NIR spectra or compressed NIR information under the form of PCA scores and MCR-ALS concentration profiles were tested as the seed information to build MSPC models. New on-line PCA-based MSPC approaches, some inspired on local rank exploratory methods for process analysis, are proposed and work as follows: a) MSPC based on individual process observation models, where multiple local PCA models are built considering the sole information in each observation point; b) Fixed Size Moving Window - MSPC, in which local PCA models are built considering a moving window of the current and few past observation points; and c) Evolving MSPC, where local PCA models are built with an increasing window of observations covering all points since the beginning of the process until the current observation. Performance of different approaches has been assessed in terms of sensitivity to fault detection and number of false alarms. The outcome of this work will be of general use to define strategies for on-line process monitoring and control and, in a more specific way, to improve quality control of petroleum derived fuels and other substances submitted

  18. Generator Set Durability Testing Using 25% ATJ Fuel Blend

    Science.gov (United States)

    2016-02-01

    UNCLASSIFIED UNCLASSIFIED 3 Table 1. Chemical & Physical Properties of Evaluated 25% ATJ Blend Test ASTM Method Units SwRI Sample ID...25% ATJ Blend Test ASTM Method Units SwRI Sample ID CL15-8613 Results Min Max Flash Point D93 °C 56.5 38 Density D4052 Test...Chemical & Physical Properties of Evaluated 25% ATJ Blend Test ASTM Method Units SwRI Sample ID CL15-8613 Results Min Max Fuel System Icing Inhibitor

  19. Used fuel rail shock and vibration testing options analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Klymyshyn, Nicholas A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-25

    The objective of the rail shock and vibration tests is to complete the framework needed to quantify loads of fuel assembly components that are necessary to guide materials research and establish a technical basis for review organizations such as the U.S. Nuclear Regulatory Commission (NRC). A significant body of experimental and numerical modeling data exists to quantify loads and failure limits applicable to normal conditions of transport (NCT) rail transport, but the data are based on assumptions that can only be verified through experimental testing. The test options presented in this report represent possible paths for acquiring the data that are needed to confirm the assumptions of previous work, validate modeling methods that will be needed for evaluating transported fuel on a case-by-case basis, and inform material test campaigns on the anticipated range of fuel loading. The ultimate goal of this testing is to close all of the existing knowledge gaps related to the loading of used fuel under NCT conditions and inform the experiments and analysis program on specific endpoints for their research. The options include tests that would use an actual railcar, surrogate assemblies, and real or simulated rail transportation casks. The railcar carrying the cradle, cask, and surrogate fuel assembly payload would be moved in a train operating over rail track modified or selected to impart shock and vibration forces that occur during normal rail transportation. Computer modeling would be used to help design surrogates that may be needed for a rail cask, a cask’s internal basket, and a transport cradle. The objective of the design of surrogate components would be to provide a test platform that effectively simulates responses to rail shock and vibration loads that would be exhibited by state-of-the-art rail cask, basket, and/or cradle structures. The computer models would also be used to help determine the placement of instrumentation (accelerometers and strain gauges

  20. Destructive distillation under pressure

    Energy Technology Data Exchange (ETDEWEB)

    1932-09-08

    A process of destructive distillation of distillable carbonaceous material under pressure is described, consisting of regulating the temperature by introducing the carbonaceous materials to a point where the reaction of hydrogenation has begun but has not stopped, by placing it in indirect heat-exchange with a cooling agent at a critical temperature below the reaction temperature, the agent being under pressure and introduced in the liquid state. Water is used as the cooling agent.

  1. Hydrogen Fuel Cell Vehicle Fuel Economy Testing at the U.S. EPA National Vehicle and Fuel Emissions Laboratory (SAE Paper 2004-01-2900)

    Science.gov (United States)

    The introduction of hydrogen fuel cell vehicles and their new technology has created the need for development of new fuel economy test procedures and safety procedures during testing. The United States Environmental Protection Agency-National Vehicle Fuels and Emissions Laborato...

  2. Modal testing and identification of a PWR fuel assembly

    International Nuclear Information System (INIS)

    Pisapia, S.; Collard, B.; Mori, V.; Bellizzi, S.

    2003-01-01

    This study aims at characterizing the vibratory behavior of a full-scale fuel assembly using an experimental approach. The effect of the assembly environment (air, stagnant water, and water under flow) is studied. The analysis of the test series shows that the vibratory behavior of full-scale fuel assembly is strongly nonlinear. An identification phase, based on temporal mean square criterion, allows us to obtain a nonlinear model representative of the first vibration mode of a fuel assembly. The selected class of models including damping and stiffness nonlinear terms is efficient in air, in stagnant water, and in water under flow. In all environments, the stiffness decreases with the displacement level and the damping increases with the velocity level. In the presence of water, the damping goes up and increases again with flowrate. (author)

  3. Test requirement for PIE of HANARO irradiated fuel rod

    International Nuclear Information System (INIS)

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

    2000-06-01

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

  4. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  5. Thermomechanical modeling of the Spent Fuel Test-Climax

    International Nuclear Information System (INIS)

    Butkovich, T.R.; Patrick, W.C.

    1986-02-01

    The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C

  6. Thermomechanical modeling of the Spent Fuel Test-Climax

    Energy Technology Data Exchange (ETDEWEB)

    Butkovich, T.R.; Patrick, W.C.

    1986-02-01

    The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C.

  7. ETHANOL DEHYDRATION IN PACKED DISTILLATION COLUMN USING GLYCEROL AS ENTRAINER: EXPERIMENTS AND HETP EVALUATION

    OpenAIRE

    Souza,W. L. R.; Silva,C. S.; Meleiro,L. A. C.; Mendes,M. F.

    2016-01-01

    Abstract The ethanol-water separation is very important because ethanol is widely applied in the chemical industry and its use as a fuel can reduce the pollution emitted to the air. However, anhydrous ethanol production using conventional distillation is impossible, at atmospheric pressure, due to the presence of an azeotrope. In the present work, experimental tests were carried out in order to evaluate the use of glycerol as an entrainer, in substitution of ethylene glycol in an extractive d...

  8. Microbiological test results of the environmental control and life support systems vapors compression distillation subsystem recycle tank components following various pretreatment protocols

    Science.gov (United States)

    Huff, Tim

    1993-01-01

    Microbiological samples were collected from the recycle tank of the vapor compression distillation (VCD) subsystem of the water recovery test at NASA MSFC following a 68-day run. The recycle tank collects rejected urine brine that was pretreated with a commercially available oxidant (Oxone) and sulfuric acid and pumps it back to the processing component of the VCD. Samples collected included a water sample and two swab samples, one from the particulate filter surface and a second from material floating on the surface of the water. No bacteria were recovered from the water sample. Both swab samples contained a spore-forming bacterium, Bacillus insolitus. A filamentous fungus was isolated from the floating material. Approximately 1 month after the pretreatment chemicals were changed to sodium hypochlorite and sulfuric acid, a swab of the particulate filter was again analyzed for microbial content. One fungus was isolated, and spore-forming bacteria were observed. These results indicate the inability of these pretreatments to inhibit surface attachment. The implications of the presence of these organisms are discussed.

  9. Vibration test and endurance test for HANARO 36-element fuel assembly

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Kim, Heon ll; Chung, Heung June

    1998-06-01

    Vibration test and endurance test for HANARO DU (depleted uranium) 36-element fuel assembly which was fabricated by KAERI were carried out based on the HANARO operation conditions. The endurance test of 22 days was added to the previous 18 days test. The vibration test was performed at various flow rates. Vibration frequency for the 36-element fuel assembly is between 11 to 14.5 Hz. And the maximum vibration displacement is less than 100 μm. From the endurance test result, it can be concluded that the appreciable fretting wear for the 36-element fuel assembly and the hexagonal flow tube was not observed. (author). 4 refs., 5 tabs., 29 figs

  10. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

  11. Foucault current testing of ferritic steel fuel cans

    International Nuclear Information System (INIS)

    Stossel, A.

    1984-10-01

    The analysis of impedance involved by a Foucault current test of ferritic steel tubes, is quite different from the classical analysis which refers to non-magnetic tubes; more particularly, volume defects are considered as magnetic anomalies. Contrarily to current instructions which recommend to test the product in a satured magnetic state, it is very interesting to work with a continuous energizing field, comparatively low, corresponding to a sequenced magnetization, of which value is obtained according to the magnetic structure of the product. This analysis is useful when testing fast reactor fuel cans [fr

  12. Improved Accelerated Stress Tests Based on Fuel Cell Vehicle Data

    Energy Technology Data Exchange (ETDEWEB)

    Patterson, Timothy [Research Engineer; Motupally, Sathya [Research Engineer

    2012-06-01

    UTC will led a top-tier team of industry and national laboratory participants to update and improve DOE’s Accelerated Stress Tests (AST’s) for hydrogen fuel cells. This in-depth investigation will focused on critical fuel cell components (e.g. membrane electrode assemblies - MEA) whose durability represented barriers for widespread commercialization of hydrogen fuel cell technology. UTC had access to MEA materials that had accrued significant load time under real-world conditions in PureMotion® 120 power plant used in transit buses. These materials are referred to as end-of-life (EOL) components in the rest of this document. Advanced characterization techniques were used to evaluate degradation mode progress using these critical cell components extracted from both bus power plants and corresponding materials tested using the DOE AST’s. These techniques were applied to samples at beginning-of-life (BOL) to serve as a baseline. These comparisons advised the progress of the various failure modes that these critical components were subjected to, such as membrane degradation, catalyst support corrosion, platinum group metal dissolution, and others. Gaps in the existing ASTs predicted the degradation observed in the field in terms of these modes were outlined. Using the gaps, new AST’s were recommended and tested to better reflect the degradation modes seen in field operation. Also, BOL components were degraded in a test vehicle at UTC designed to accelerate the bus field operation.

  13. Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary

    International Nuclear Information System (INIS)

    Gregson, Michael Warren; Brockmann, John E.; Nolte, O.; Loiseau, O.; Koch, W.; Molecke, Martin Alan; Autrusson, Bruno; Pretzsch, Gunter Guido; Billone, M. C.; Lucero, Daniel A.; Burtseva, T.; Brucher, W; Steyskal, Michele D.

    2006-01-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO 2 , test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments

  14. 40 CFR 60.4325 - What emission limits must I meet for NOX if my turbine burns both natural gas and distillate oil...

    Science.gov (United States)

    2010-07-01

    ... NOX if my turbine burns both natural gas and distillate oil (or some other combination of fuels)? 60... both natural gas and distillate oil (or some other combination of fuels)? You must meet the emission... burning that fuel. Similarly, when your total heat input is greater than 50 percent distillate oil and...

  15. 40 CFR 86.340-79 - Gasoline-fueled engine dynamometer test run.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Gasoline-fueled engine dynamometer... Emission Regulations for New Gasoline-Fueled and Diesel-Fueled Heavy-Duty Engines; Gaseous Exhaust Test Procedures § 86.340-79 Gasoline-fueled engine dynamometer test run. (a) This section applies to gasoline...

  16. 40 CFR 1051.515 - How do I test my fuel tank for permeation emissions?

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false How do I test my fuel tank for... Procedures § 1051.515 How do I test my fuel tank for permeation emissions? Measure permeation emissions by weighing a sealed fuel tank before and after a temperature-controlled soak. (a) Preconditioning fuel soak...

  17. Development of 3-Pin Fuel Test Loop and Utilization Technology

    International Nuclear Information System (INIS)

    Lee, Chung Young; Sim, B. S.; Lee, C. Y.

    2007-06-01

    The principal contents of this project are to design, fabricate and install the steady-state fuel test loop in HANARO for nuclear technology development. Procurement and, fabrication of main equipment, licensing and installation for fuel test loop have been performed. Following contents are described in the report. 1. Design - Design of the In-pile system and Out pile system 2. Fabrication and procurement of the equipment - Fabrication of the In-pile system and In-pool piping - Fabrication and procurement of the equipment of the out-pile system 3. Acquisition of the license - Preparation of the safety analysis report and acquisition of the license - Pre-service inspection of the facility 4. Installation and commissioning - Installation of the FTL - Development of the commissioning procedure

  18. ORR irradiation experiment OF-1: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

    1977-08-01

    The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10 22 neutrons/cm 2 (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure

  19. Melcor benchmarking against integral severe fuel damage tests

    Energy Technology Data Exchange (ETDEWEB)

    Madni, I.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-09-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the U.S. Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC to provide independent assessment of MELCOR, and a very important part of this program is to benchmark MELCOR against experimental data from integral severe fuel damage tests and predictions of that data from more mechanistic codes such as SCDAP or SCDAP/RELAP5. Benchmarking analyses with MELCOR have been carried out at BNL for five integral severe fuel damage tests, namely, PBF SFD 1-1, SFD 14, and NRU FLHT-2, analyses, and their role in identifying areas of modeling strengths and weaknesses in MELCOR.

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

    Science.gov (United States)

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

    2011-01-01

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

  1. Modular, High-Volume Fuel Cell Leak-Test Suite and Process

    Energy Technology Data Exchange (ETDEWEB)

    Ru Chen; Ian Kaye

    2012-03-12

    Fuel cell stacks are typically hand-assembled and tested. As a result the manufacturing process is labor-intensive and time-consuming. The fluid leakage in fuel cell stacks may reduce fuel cell performance, damage fuel cell stack, or even cause fire and become a safety hazard. Leak check is a critical step in the fuel cell stack manufacturing. The fuel cell industry is in need of fuel cell leak-test processes and equipment that is automatic, robust, and high throughput. The equipment should reduce fuel cell manufacturing cost.

  2. Accident situations tests HTR fuel with the device Kufa

    International Nuclear Information System (INIS)

    Kellerbauer, A. I.; Freis, D.

    2010-01-01

    The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

  3. Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1995-03-01

    In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

  4. Fuel cell climatic tests designed for new configured aircraft application

    International Nuclear Information System (INIS)

    Begot, Sylvie; Harel, Fabien; Candusso, Denis; Francois, Xavier; Pera, Marie-Cecile; Yde-Andersen, Steen

    2010-01-01

    The implementation of Fuel Cell (FC) systems in transportation systems, as aircrafts, requires some better understanding and mastering of the new generator behaviours in low temperature environments. To this end, a PEMFC stack is tested and characterised in a climatic chamber. The impacts of the low temperatures over different FC operation and start-up conditions are estimated using a specific test bench developed in-lab. Some descriptions concerning the test facilities and the experimental set-up are given in the paper, as well as some information about the test procedures applied. Some examples of test results are shown and analysed. The experiments are derived from aircraft requirements and are related with different scenarios of airplane operation. Finally, some assessments concerning the FC system behaviour in low temperature conditions are made, especially with regard to the constraints to be encountered by the next embedded FC generators.

  5. Fuel cell climatic tests designed for new configured aircraft application

    Energy Technology Data Exchange (ETDEWEB)

    Begot, Sylvie; Pera, Marie-Cecile [FC LAB, Rue Thierry Mieg, F 90010 Belfort Cedex (France); Franche-Comte Electronique Mecanique Thermique et Optique - Sciences et Technologies (FEMTO-ST), Departement energie et ingenierie des systemes multiphysiques (ENISYS), Unite Mixte de Recherche (UMR) du Centre National de la Recherche Scientifique (CNRS) 6174, University of Franche-Comte (UFC) (France); Harel, Fabien; Candusso, Denis [FC LAB, Rue Thierry Mieg, F 90010 Belfort Cedex (France); The French National Institute for Transport and Safety Research (INRETS), Transports and Environment Laboratory (LTE), Laboratory for New Technologies (LTN) (France); Francois, Xavier [FC LAB, Rue Thierry Mieg, F 90010 Belfort Cedex (France); FC LAB, University of Technology Belfort-Montbeliard (UTBM) (France); Yde-Andersen, Steen [IRD Fuel Cells A/S, Kullinggade 31, 5700 Svendborg (Denmark)

    2010-07-15

    The implementation of Fuel Cell (FC) systems in transportation systems, as aircrafts, requires some better understanding and mastering of the new generator behaviours in low temperature environments. To this end, a PEMFC stack is tested and characterised in a climatic chamber. The impacts of the low temperatures over different FC operation and start-up conditions are estimated using a specific test bench developed in-lab. Some descriptions concerning the test facilities and the experimental set-up are given in the paper, as well as some information about the test procedures applied. Some examples of test results are shown and analysed. The experiments are derived from aircraft requirements and are related with different scenarios of airplane operation. Finally, some assessments concerning the FC system behaviour in low temperature conditions are made, especially with regard to the constraints to be encountered by the next embedded FC generators. (author)

  6. Destructive distillation: oils

    Energy Technology Data Exchange (ETDEWEB)

    West, J; Glover, S

    1918-01-31

    Canned and other coals are destructively distilled in continuously operated vertical retorts which at their upper portions are maintained at temperatures suitable for low temperature oil distillation such as about 700/sup 0/C, and at their lower portions the temperature is higher and such as to be suitable for the production of gas, e.g., about 1400/sup 0/C. Superheated steam is introduced into the lower portion of the retort, preferably by means of the arrangement described in Specification 120,458, and this is converted into blue water gas which assists the distillation in the center of the coal charge. The retorts are preferably such as are described in Specifications 2663/07 and 7757/14.

  7. Distilling hydrocarbons, etc

    Energy Technology Data Exchange (ETDEWEB)

    Turner, C

    1917-11-23

    In the fractional or destructive distillation of hydrocarbon oils or other liquids, the pressure in the still is raised and lowered alternately. The still is closed to raise the pressure, and is opened to lower the pressure rapidly solely by expansion of the vapors. The operation is effected without intermittent cooling, except such as may occur during the lowering of the pressure. In distilling hydrocarbon oil, pressure steam is blown into the oil until the pressure reaches 5 lb/in./sup 2/. The vapor outlet is then opened until the pressure falls to 2 lb/in./sup 2/, whereupon the vapor outlet is closed and steam is again admitted. The operation is continued until the steam, which is of 20 lb pressure, no longer effects distillation; after this stage, superheated steam is used.

  8. Comparative tests of bench equipment for fuel control system testing of gas-turbine engine

    Science.gov (United States)

    Shendaleva, E. V.

    2018-04-01

    The relevance of interlaboratory comparative researches is confirmed by attention of world metrological community to this field of activity. Use of the interlaboratory comparative research methodology not only for single gages collation, but also for bench equipment complexes, such as modeling stands for fuel control system testing of gas-turbine engine, is offered. In this case a comparative measure of different bench equipment will be the control fuel pump. Ensuring traceability of measuring result received at test benches of various air enterprises, development and introduction of national standards to practice of bench tests and, eventually, improvement of quality and safety of a aircraft equipment is result of this approach.

  9. Fractional distillation of oil

    Energy Technology Data Exchange (ETDEWEB)

    Jones, L D

    1931-10-31

    A method of dividing oil into lubricating oil fractions without substantial cracking by introducing the oil in a heated state into a fractionating column from which oil fractions having different boiling points are withdrawn at different levels, while reflux liquid is supplied to the top of the column, and additional heat is introduced into the column by contacting with the oil therein a heated fluid of higher monlecular weight than water and less susceptible to thermal decomposition than is the highest boiling oil fraction resulting from the distillation, or of which any products produced by thermal decomposition will not occur in the highest boiling distillate withdrawn from the column.

  10. Distilling carbonaceous materials

    Energy Technology Data Exchange (ETDEWEB)

    Trumble, M J

    1925-06-29

    Carbonaceous materials such as coal, oil shale, peat, or wood are destructively distilled while being subjected to the action of superheated steam and hydrogen, the latter being provided by dissociating a part of the superheated steam. The materials are charged into a retort heated by a burner and superheated steam and hydrogen are passed in by a pipe and nozzles. The distillates enter a dust extractor through openings and escape through openings shielded by cones into an outlet pipe leading to condensers. The dust which settles in the bottom of the apparatus is periodically removed.

  11. Distilling coal, etc

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, J

    1917-12-21

    Coals of various kinds such as shales, bitumens, and oil sand, peat, etc. are distilled at 350 to 450/sup 0/C and in the presence of vapors and gases obtained by cracking hydrocarbon oils, or the gases obtained by separating the condensable hydrocarbons therefrom, and, if desired, with the addition of superheated steam. The hydrocarbons are properly cracked by passing through molten lead as described in Specification 116,304. According to the Provisional Specification, superheated steam alone may be used to effect the distillation.

  12. Underwater Coatings Testing for INEEL Fuel Basin Applications

    International Nuclear Information System (INIS)

    Julia L. Tripp

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is deactivating several fuel storage basins. Airborne contamination is a concern when the sides of the basins are exposed and allowed to dry during water removal. One way of controlling this airborne contamination is to fix the contamination in place while the pool walls are still submerged. There are many underwater coatings available on the market that are used in marine, naval and other applications. A series of tests were run to determine whether the candidate underwater fixatives are easily applied and adhere well to the substrates (pool wall materials) found in INEEL fuel pools. The four pools considered included (1) Test Area North (TAN-607) with epoxy painted concrete walls; (2) Idaho Nuclear Technology and Engineering Center (INTEC) (CPP-603) with bare concrete walls; (3) Materials Test Reactor (MTR) Canal with stainless steel lined concrete walls; and (4) Power Burst Facility (PBF-620) with stainless steel lined concrete walls on the bottom and epoxy painted carbon steel lined walls on the upper portions. Therefore, the four materials chosen for testing included bare concrete, epoxy painted concrete, epoxy painted carbon steel, and stainless steel. The typical water temperature of the pools varies from 55 F to 80 F dependent on the pool and the season. These tests were done at room temperature

  13. Underwater Coatings Testing for INEEL Fuel Basin Applications

    Energy Technology Data Exchange (ETDEWEB)

    Julia L. Tripp

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is deactivating several fuel storage basins. Airborne contamination is a concern when the sides of the basins are exposed and allowed to dry during water removal. One way of controlling this airborne contamination is to fix the contamination in place while the pool walls are still submerged. There are many underwater coatings available on the market that are used in marine, naval and other applications. A series of tests were run to determine whether the candidate underwater fixatives are easily applied and adhere well to the substrates (pool wall materials) found in INEEL fuel pools. The four pools considered included (1) Test Area North (TAN-607) with epoxy painted concrete walls; (2) Idaho Nuclear Technology and Engineering Center (INTEC) (CPP-603) with bare concrete walls; (3) Materials Test Reactor (MTR) Canal with stainless steel lined concrete walls; and (4) Power Burst Facility (PBF-620) with stainless steel lined concrete walls on the bottom and epoxy painted carbon steel lined walls on the upper portions. Therefore, the four materials chosen for testing included bare concrete, epoxy painted concrete, epoxy painted carbon steel, and stainless steel. The typical water temperature of the pools varies from 55 F to 80 F dependent on the pool and the season. These tests were done at room temperature.

  14. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2005-06-01

    The disposal canister for spent nuclear fuel will be transferred by a lift to the repository, which is 500 m deep in the bedrock. Model tests were carried out with the objective to estimate weather feasible shock absorber can be developed against the design accident case where the canister should survive a free fall to the lift shaft. If the velocity of the canister is not controlled by air drag or by any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity in impact on water when the bottom pit of the lift well is filled with groundwater. However, the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20 m high filling to the bottom pit of the lift well by Light Expanded Clay Aggregate (LECA), gives fair impact absorption to protect the fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

  15. Tests of prototype salt stripper system for IFR fuel cycle

    International Nuclear Information System (INIS)

    Carls, E.L.; Blaskovitz, R.J.; Johnson, T.R.; Ogata, T.

    1993-01-01

    One of the waste treatment steps for the on-site reprocessing of spent fuel from the Integral Fast Reactor fuel cycles is stripping of the electrolyte salt used in the electrorefining process. This involves the chemical reduction of the actinides and rare earth chlorides forming metals which then dissolve in a cadmium pool. To develop the equipment for this step, a prototype salt stripper system has been installed in an engineering scale argon-filled glovebox. Pumping trails were successful in transferring 90 kg of LiCl-KCl salt containing uranium and rare earth metal chlorides at 500 degree C from an electrorefiner to the stripper vessel at a pumping rate of about 5 L/min. The freeze seal solder connectors which were used to join sections of the pump and transfer line performed well. Stripping tests have commenced employing an inverted cup charging device to introduce a Cd-15 wt % Li alloy reductant to the stripper vessel

  16. Certification testing of the MOX Fresh Fuel Package (MFFP)

    International Nuclear Information System (INIS)

    Nichols, J.C. III; Yapuncich, F.L.

    2004-01-01

    Packaging Technology, Inc. (PacTec) is designing the MFFP as part of the Duke, COGEMA, Stone and Webster (DCS) consortium. DCS is tasked with providing the Department of Energy (DOE) with domestic MOX fuel fabrication and reactor irradiation services for the purpose of disposing of surplus weapons usable plutonium. This paper will discuss the development of the MFFP certification test program. The MFFP was subjected to a total of eleven free and puncture drops of the course of the certification testing. Because of the plutonium content, the design must be a Type BF, which among other things requires a containment boundary with a tested leakage rate of 1 x 10 -7 cm 3 /s air at 1 atm absolute and 25 C, or less. Both economics (desire for maximized payload) and operational (conveyance mode restricts size and weight) constraints lead to a highly optimized design. The optimized package design led to a significant test program which needed to address the containment boundary stability, puncture resistance of the package and lid end impact limiter, structural performance of the light weight lid structure, and stability of the internal structures. The test program efficiently balanced the test objectives while minimizing the number of costly hardware items used during this destructive testing. This balance achieved by strategic replacement of mock and prototypic payloads, impact limiters, and by careful test order considerations. The paper will conclude with a selected summary of the testing and an assessment of the test programs thoroughness

  17. Focused ultrasonic wave testing, in immersion of spent fuel cans

    International Nuclear Information System (INIS)

    Poinboeuf, P.; Furlan, J.

    1984-10-01

    To detect weak and very weak damage of the fuel can, ultrasonic testing has been used. For that, a simple mechanical device, allowing to maintain an optimal ultrasonic focussing on irradiated cans, is presented. Its aim is to correct the variation of the incidence angle due to the possible ovalization of pins. After a description of the device, the results obtained with tests carried out on non-irradiated cans, including artificial ovalized regions, standard defects, are presented. After the description of the adaptation of this mechanism on a test bench which allows an helicoidal exploration of pins, some results obtained in hot cell during examinations experimental pins and previously tested by Foucault current [fr

  18. Ignition delay times of Gasoline Distillation Cuts measured with Ignition Quality Tester

    KAUST Repository

    Naser, Nimal; Singh, Eshan; Ahmed, Ahfaz; Sarathy, Mani

    2017-01-01

    Tailoring fuel properties to maximize the efficiency of internal combustion engines is a way towards achieving cleaner combustion systems. In this work, the ignition properties of various gasoline fuel distillation cuts are analyzed to better

  19. Design verification test of instrumented capsule (02F-11K) for nuclear fuel irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, J. M.; Oh, J. M. [and others

    2004-01-01

    An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (Self-Powered Neutron Detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. The test fuel rods were irradiated at less than 350 W/cm to 5.13 GWD/MTU with fuel centerline peak temperature below 1,375 .deg. C. The structural stability of the capsule was satisfied by the naked eye in service pool of HANARO. The capsule and test fuel rods were dismantled and test fuel rods were examined at the hot cell of IMEF (Irradiated Material Examination Facility)

  20. The Third Dryout Fuel Behaviour Test Series in IFA-613

    International Nuclear Information System (INIS)

    Ianiri, Raffaella

    1998-02-01

    The objective of the dryout experiment with the instrumented fuel assembly IFA-613 is to provide information on the consequences induced on fuel by short terms dry outs having characteristics similar to those anticipated to occur from pump trips in a Boiling Water Reactor (BWR). For the third experiment it was planned to test one fresh and two pre-irradiated segments. Unfortunately one of the channels, Channel A developed a leakage and was not suitable for testing anymore. The rig was loaded with only two rods: one fresh PWR rod with a design similar to the fresh rod in IFA-613.1 and one pre-irradiated PWR segment (N1310 with a burn-up of 29 MWd/kgU). Both rods were equipped with a clad extensometer and two clad surface thermocouples (upper and lower position). The rig was loaded during the December 1997 shutdown and the dryout tests were performed on 16th January 1998. Both rods experienced temperature excursions with a target peak clad temperature (PCT) of 650 o C. According to the measured cladding temperatures, the time above the target temperature was about 4-5 s for both rods. The lower thermocouple did not indicate dryout at any occasion. The rig was unloaded immediately after the testing. (author)

  1. 40 CFR 1060.520 - How do I test fuel tanks for permeation emissions?

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false How do I test fuel tanks for... STATIONARY EQUIPMENT Test Procedures § 1060.520 How do I test fuel tanks for permeation emissions? Measure permeation emissions by weighing a sealed fuel tank before and after a temperature-controlled soak. (a...

  2. 33 CFR 183.580 - Static pressure test for fuel tanks.

    Science.gov (United States)

    2010-07-01

    ... pressure test for fuel tanks. A fuel tank is tested by performing the following procedures in the following... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Static pressure test for fuel tanks. 183.580 Section 183.580 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND...

  3. Distillation Column Flooding Predictor

    Energy Technology Data Exchange (ETDEWEB)

    George E. Dzyacky

    2010-11-23

    The Flooding Predictor™ is a patented advanced control technology proven in research at the Separations Research Program, University of Texas at Austin, to increase distillation column throughput by over 6%, while also increasing energy efficiency by 10%. The research was conducted under a U. S. Department of Energy Cooperative Agreement awarded to George Dzyacky of 2ndpoint, LLC. The Flooding Predictor™ works by detecting the incipient flood point and controlling the column closer to its actual hydraulic limit than historical practices have allowed. Further, the technology uses existing column instrumentation, meaning no additional refining infrastructure is required. Refiners often push distillation columns to maximize throughput, improve separation, or simply to achieve day-to-day optimization. Attempting to achieve such operating objectives is a tricky undertaking that can result in flooding. Operators and advanced control strategies alike rely on the conventional use of delta-pressure instrumentation to approximate the column’s approach to flood. But column delta-pressure is more an inference of the column’s approach to flood than it is an actual measurement of it. As a consequence, delta pressure limits are established conservatively in order to operate in a regime where the column is never expected to flood. As a result, there is much “left on the table” when operating in such a regime, i.e. the capacity difference between controlling the column to an upper delta-pressure limit and controlling it to the actual hydraulic limit. The Flooding Predictor™, an innovative pattern recognition technology, controls columns at their actual hydraulic limit, which research shows leads to a throughput increase of over 6%. Controlling closer to the hydraulic limit also permits operation in a sweet spot of increased energy-efficiency. In this region of increased column loading, the Flooding Predictor is able to exploit the benefits of higher liquid

  4. Vacuum distilling vessel

    Energy Technology Data Exchange (ETDEWEB)

    Reik, H

    1928-12-27

    Vacuum distilling vessel for mineral oil and the like, characterized by the ring-form or polyconal stiffeners arranged inside, suitably eccentric to the casing, being held at a distance from the casing by connecting members of such a height that in the resulting space if necessary can be arranged vapor-distributing pipes and a complete removal of the residue is possible.

  5. Distilling bituminous materials

    Energy Technology Data Exchange (ETDEWEB)

    Forrest, C N; Hayden, H P

    1926-12-28

    Bituminous materials such as heavy asphaltic residues from petroleum distillation or natural bitumens such as gilsonite, grahamite, etc. are distilled in presence of a carrier consisting of pieces of inert refractory material and by the heat generated by combustion of the coke which remains on the carrier after distillation. A vertical cylindrical retort, in which a wood and coal fire is first made, is charged with an inert refractory substance such as pumice stone, broken firebrick, burnt fire clay, carborundum, ash, etc. mixed with a bituminous substance, which, if fusible, may be melted and added in a rotary drum. The mixture passes downwards through the retort, first through regions in which the hydrocarbons are distilled and cracked and then through a region in which the remaining carbon is burnt by a limited supply of air and steam admitted through a grate. The inert material is discharged through a water seal and used again. Vapors, withdrawn from the retort though an outlet, pass through a heat intercharger and separator and are treated with a spray of sulfuric acid to separate ammonia in a scrubber, with water sprays to condense oil in scrubbers, and with oil in a scrubber.

  6. Analytical determination of distillation boundaries for ternary azeotropic systems

    OpenAIRE

    Marcilla Gomis, Antonio; Reyes Labarta, Juan Antonio; Velasco, Raúl; Serrano Cayuelas, María Dolores; Olaya López, María del Mar

    2009-01-01

    A new straight forward algorithm to calculate distillation boundaries in ternary azeotropic systems has been developed. The proposed method allows, using cubic splines, the calculation of distillation trajectories and the calculation of that corresponding to the searched distillation boundaries. The algorithm is applied to 4 ternary liquid-vapour systems to test its validity. Vicepresidency of Research (University of Alicante) and Generalitat Valenciana (GV/2007/125)

  7. Heating oil, distillates and residuals outlook

    International Nuclear Information System (INIS)

    Ervin, M.J.

    2004-01-01

    M.J. Ervin and Associates offers strategic planning support to the downstream petroleum industry in terms of price market monitoring, market analysis, media commentary and benchmarking of marketing operations. This presentation included graphs depicting supply and demand for heating oil distillates and residuals from the mid 1990s to 2004. It was noted that the long-term decline in residuals demand in the United States and Canada are due to environmental issues, the use of natural gas as an alternative, and the increasing complexity of refineries. Graphs depicting market impacts of refinery utilization and inventory trends showed that middle distillate production will increase at the expense of gasoline production. Middle distillates and gasoline markets will be more sensitive to supply disruptions, resulting in more frequent price spikes. Inventory trends indicate a greater reliance on product imports. The demand for heating fuel has stabilized due to the continued penetration of natural gas in eastern states and provinces. The demand for diesel fuel has growth 1.5 to 2 per cent while the demand for jet fuel has remained relatively flat and depends greatly on the growth of the gross national product (GNP). tabs., figs

  8. Membrane distillation for milk concentration

    NARCIS (Netherlands)

    Moejes, S.N.; Romero Guzman, Maria; Hanemaaijer, J.H.; Barrera, K.H.; Feenstra, L.; Boxtel, van A.J.B.

    2015-01-01

    Membrane distillation is an emerging technology to concentrate liquid products while producing high quality water as permeate. Application for desalination has been studied extensively the past years, but membrane distillation has also potential to produce concentrated food products like

  9. Vacuum Drying Tests for Storage of Aluminum Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Chen, K.F.; Large, W.S.; Sindelar, R.L.

    1998-05-01

    A total inventory of up to approximately 32,000 aluminum-based spent nuclear fuel (Al SNF) assemblies are expected to be shipped to Savannah River Site (SRS) from domestic and foreign research reactors over the next several decades. Treatment technologies are being developed as alternatives to processing for the ultimate disposition of Al SNF in the geologic repository. One technology, called Direct/Co-disposal of Al SNF, would place the SNF into a canister ready for disposal in a waste package, with or without canisters containing high-level radioactive waste glass logs, in the repository. The Al SNF would be transferred from wet storage and would need to be dried in the Al SNF canister. The moisture content inside the Al SNF canister is limited to avoid excessive Al SNF corrosion and hydrogen buildup during interim storage before disposal. A vacuum drying process was proposed to dry the Al SNF in a canister. There are two major concerns for the vacuum drying process. One is water inside the canister could become frozen during the vacuum drying process and the other one is the detection of dryness inside the canister. To vacuum dry an irradiated fuel in a heavily shielded canister, it would be very difficult to open the lid to inspect the dryness during the vacuum drying operation. A vacuum drying test program using a mock SNF assembly was conducted to demonstrate feasibility of drying the Al SNF in a canister. These tests also served as a check-out of the drying apparatus for future tests in which irradiated fuel would be loaded into a canister under water followed by drying for storage

  10. Fuel bundle examination techniques for the Phebus fission product test

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Clement, B.; Hardt, P. von der

    1996-01-01

    The paper develops the non-destructive examinations, with a special emphasis on transmission tomography, performed in the Phebus facility, using a linear accelerator associated with a line scan camera based on PCD components. This particular technique enabled the high level of penetration to be obtained, necessary for this high density application. Spatial resolution is not far from the theoretical limit and the density resolution is often adequate. This technique permitted: 1) to define beforehand the cuts on a precise basis, avoiding a long step-by-step choice as in previous in-pile tests; 2) to determine, at an early stage, mass balance, material relocations (in association with axial gamma spectrometry), and FP distribution, as an input into re-calculations of the bundle events. However, classical cuttings, periscopic visual examinations, macrographies, micrographies and EPMA analyses remain essential to give oxidation levels (in the less degraded zones), phase aspect and composition, to distinguish between materials of identical density, and, if possible, to estimate temperatures. Oxidation resistance of sensors (thermocouples or ultrasonic thermometers) is also traced. The EPMA gives access to the molten material chemical analyses, especially in the molten fuel blockage area. The first results show that an important part of the fuel bundle melted (which was one of the objectives of this test) and that the degradation level is close to TIMI-2 with a molten plug under a cavity surrounded by an uranium-rich crust. In lower and upper areas fuel rods are less damaged. Complementaries between these examination techniques and between international teams involved will be major advantages in the Phebus FPT0 test comprehension. 3 refs, 9 figs

  11. Liquefied Gaseous Fuels Spill Test Facility: Overview of STF capabilities

    International Nuclear Information System (INIS)

    Gray, H.E.

    1993-01-01

    The Liquefied Gaseous Fuels Spill Test Facility (STF) constructed at the Department of Energy's Nevada Test Site is a basic research tool for studying the dynamics of accidental releases of various hazardous liquids. This Facility is designed to (1) discharge, at a controlled rate, a measured volume of hazardous test liquid on a prepared surface of a dry lake bed (Frenchman Lake); (2) monitor and record process operating data, close-in and downwind meteorological data, and downwind gaseous concentration levels; and (3) provide a means to control and monitor these functions from a remote location. The STF will accommodate large and small-scale testing of hazardous test fluid release rates up to 28,000 gallons per minute. Spill volumes up to 52,800 gallons are achievable. Generic categories of fluids that can be tested are cryogenics, isothermals, aerosol-forming materials, and chemically reactive. The phenomena that can be studied include source definition, dispersion, and pool fire/vapor burning. Other capabilities available at the STF include large-scale wind tunnel testing, a small test cell for exposing personnel protective clothing, and an area for developing mitigation techniques

  12. Proton Exchange Membrane Fuel Cell Engineering Model Powerplant. Test Report: Benchmark Tests in Three Spatial Orientations

    Science.gov (United States)

    Loyselle, Patricia; Prokopius, Kevin

    2011-01-01

    Proton exchange membrane (PEM) fuel cell technology is the leading candidate to replace the aging alkaline fuel cell technology, currently used on the Shuttle, for future space missions. This test effort marks the final phase of a 5-yr development program that began under the Second Generation Reusable Launch Vehicle (RLV) Program, transitioned into the Next Generation Launch Technologies (NGLT) Program, and continued under Constellation Systems in the Exploration Technology Development Program. Initially, the engineering model (EM) powerplant was evaluated with respect to its performance as compared to acceptance tests carried out at the manufacturer. This was to determine the sensitivity of the powerplant performance to changes in test environment. In addition, a series of tests were performed with the powerplant in the original standard orientation. This report details the continuing EM benchmark test results in three spatial orientations as well as extended duration testing in the mission profile test. The results from these tests verify the applicability of PEM fuel cells for future NASA missions. The specifics of these different tests are described in the following sections.

  13. Sodium-fuel interaction: dropping experiments and subassembly test

    International Nuclear Information System (INIS)

    Holtbecker, H.; Schins, H.; Jorzik, E.; Klein, K.

    1978-01-01

    Nine dropping tests, which bring together 2 to 4 kg of molten UO 2 with 150 l sodium, showed the incoherency and non-violence of these thermal interactions. The pressures can be described by sodium incipient boiling and bubble collapse; the UO 2 fragmentation by thermal stress and bubble collapse impact forces. The mildness of the interaction is principally due to the slowness and incoherency of UO 2 fragmentation. This means that parametric models which assume instantaneous mixing and fragmentation are of no use for the interpretation of dropping experiments. One parametric model, the Caldarola Fuel Coolant Interaction Variable Mass model, is being coupled to the two dimensional time dependent hydrodynamic REXCO-H code. In a first step the coupling is applicated to a monodimensional geometry. A subassembly test is proposed to validate the model. In this test rapid mixing between UO 2 and sodium has to be obtained. Dispersed molten UO 2 fuel is obtained by flashing injected sodium drops inside a UO 2 melt. This flashing is theoretically explained and modelled as a superheat limited explosion. The measured sodium drop dwell times of two experiments are compared to results obtained from the mentioned theory, which is the basis of the Press 2 Code

  14. Review of WWER fuel and material tests in the Halden reactor

    International Nuclear Information System (INIS)

    Volkov, B.; Kolstad, E.

    2006-01-01

    A review of the tests with WWER fuels and materials conducted in HBWR over the years of cooperation with Russia is presented. The first test with old generation WWER-440 fuel and PWR specification fuel was carried out from 1995 to 1998. Some differences between these fuels regarding irradiation induced densification and pellet design as well as similar fuel thermal behaviour, swelling and FGR were revealed during the test. The data from this test are reviewed and compared with PIE recently performed to confirm the in-pile measurements. The second test was started in March 1999 with the main objective to study different modified WWER fuels also in comparison with PWR fuel. The results indicated that all these modified WWER fuels exhibit improved densification properties relative to earlier tested fuel. In-pile data on fuel densification have been analysed with respect to as fabricated fuel microstructure and can be used for verification of fuel behaviour models. Corrosion and creep tests in the Halden reactor encompass WWER cladding alloys and some results are given. Prospective WWER fuel and material tests foreseen within the frame of the joint program of OECD HRP are also presented. (authors)

  15. The design of in-pile test section for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. N.; Lee, J. M.; Shim, B. S.; Zee, D. Y.; Park, S. H.; Ahn, S. H.; Lee, J. Y.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    As an equipment for nuclear fuel's general performance irradiation test in HANARO, Fuel Test Loop(FTL) has been developed that can irradiate the pin to the maximum number of 3 at the core irradiation hole(IR1 hole) by considering for it's utility and user's irradiation requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). IPS consists for IPS Vessel assembly, In-Pool Piping, IPS Support, In-Pool Piping Support etc. Design that such IPS considers interference item consisted to do not bear in existing facilities by one. IVA that is connected to the OPS are controlled and regulated by means of system pressure, system temperature and the water quality. IPS Vessel assembly is consisted of outer pressure vessel, inner pressure vessel, IPS head, inner assembly and test fuel carrier. After 3-Pin FTL development which is expected to be finished by the 2006, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

  16. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    International Nuclear Information System (INIS)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

  17. Distillation with labelled transition systems

    DEFF Research Database (Denmark)

    Hamilton, Geoffrey William; Jones, Neil

    2012-01-01

    In this paper, we provide an improved basis for the " distillation" program transformation. It is known that superlinear speedups can be obtained using distillation, but cannot be obtained by other earlier automatic program transformation techniques such as deforestation, positive supercompilation...... and partial evaluation. We give distillation an improved semantic basis, and explain how superlinear speedups can occur....

  18. Reactivity initiated accident test series Test RIA 1-4 fuel behavior report

    International Nuclear Information System (INIS)

    Cook, B.A.; Martinson, Z.R.

    1984-09-01

    This report presents and discusses results from the final test in the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-4, conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Nine preirradiated fuel rods in a 3 x 3 bundle configuration were subjected to a power burst while at boiling water reactor hot-startup system conditions. The test resulted in estimated axial peak, radial average fuel enthalpies of 234 cal/g UO 2 on the center rod, 255 cal/g UO 2 on the side rods, and 277 cal/g UO 2 on the corner rods. Test RIA 1-4 was conducted to investigate fuel coolability and channel blockage within a bundle of preirradiated rods near the present enthalpy limit of 280 cal/g UO 2 established by the US Nuclear Regulatory Commission. The test design and conduct are described, and the bundle and individual rod thermal and mechanical responses are evaluated. Conclusions from this final test and the entire PBF RIA Test Series are presented

  19. An inspection standard of fuel for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao; Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

  20. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2003-01-01

    The holding canister for spent nuclear fuel will be transferred by a lift to the final disposal tunnels 500m deep in the bedrock. Model tests were carried out with an objective to estimate weather feasible shock absorbing properties can be met in a design accident case where the canister should survive a free fall due to e.g. sabotage. If the velocity of the canister is not controlled by air drag or any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity when stricken by the surface penetration impact if the bottom pit of the lift well would be filled with groundwater. However the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20m high filling to the bottom pit of the lift well by ceramic gravel, trade mark LECA-sora, gives a fair impact absorption to protect the spent fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

  1. Autonomy-Enabled Fuel Savings for Military Vehicles: Report on 2016 Aberdeen Test Center Testing

    Energy Technology Data Exchange (ETDEWEB)

    Ragatz, Adam [National Renewable Energy Lab. (NREL), Golden, CO (United States); Prohaska, Robert [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gonder, Jeff [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2017-05-26

    Fuel savings have never been the primary focus for autonomy-enabled military vehicles. However, studies have estimated that autonomy in passenger and commercial vehicles could improve fuel economy by as much as 22%-33% over various drive cycles. If even a fraction of this saving could be realized in military vehicles, significant cost savings could be realized each year through reduced fuel transport missions, reduced fuel purchases, less maintenance, fewer required personnel, and increased vehicle range. Researchers from the National Renewable Energy Laboratory installed advanced data logging equipment and instrumentation on two autonomy-enabled convoy vehicles configured with Lockheed Martin's Autonomous Mobility Applique System to determine system performance and improve on the overall vehicle control strategies of the vehicles. Initial test results from testing conducted at the U.S. Army Aberdeen Test Center at the Aberdeen Proving Grounds are included in this report. Lessons learned from in-use testing and performance results have been provided to the project partners for continued system refinement.

  2. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  3. Method of controlling a distillatory coulumn in a liquefaction and distillation device for radioactive rare gases

    International Nuclear Information System (INIS)

    Kimura, Shigeru; Awata, Yoshihisa.

    1975-01-01

    Object: To automatically and securely perform controlling purity of column bottom liquid, which is most difficult in the liquefaction and distillation device. Structure: In a liquefaction and distillation device for liquefying and separating rare gas krypton in fuel reprocessing gases, a difference in temperature between the column bottom liquid (column top) and the distillation portion is detected so as to maintain temperature in the distillation portion of the distilling column constant, and the vaporization amount of column bottom liquid is varied with the difference in temperature to control purity of the column bottom liquid. (Kamimura, M.)

  4. Vapor compression distiller and membrane technology for water revitalization

    Science.gov (United States)

    Ashida, A.; Mitani, K.; Ebara, K.; Kurokawa, H.; Sawada, I.; Kashiwagi, H.; Tsuji, T.; Hayashi, S.; Otsubo, K.; Nitta, K.

    1987-01-01

    Water revitalization for a space station can consist of membrane filtration processes and a distillation process. Water recycling equipment using membrane filtration processes was manufactured for ground testing. It was assembled using commercially available components. Two systems for the distillation are studied: one is absorption type thermopervaporation cell and the other is a vapor compression distiller. Absorption type thermopervaporation, able to easily produce condensed water under zero gravity, was investigated experimentally and through simulated calculation. The vapor compression distiller was studied experimentally and it offers significant energy savings for evaporation of water.

  5. Fracture mapping at the Spent Fuel Test-Climax

    International Nuclear Information System (INIS)

    Wilder, D.G.; Yow, J.L. Jr.

    1981-05-01

    Mapping of geologic discontinuities has been done in several phases at the Spent Fuel Test-Climax (SFT-C) in the granitic Climax stock at the Nevada Test Site. Mapping was carried out in the tail drift, access drift, canister drift, heater drifts, instrumentation alcove, and receiving room. The fractures mapped as intersecting a horizontal datum in the canister and heater drifts are shown on one figure. Fracture sketch maps have been compiled as additional figures. Geologic mapping efforts were scheduled around and significantly impacted by the excavation and construction schedules. Several people were involved in the mapping, and over 2500 geologic discontinuities were mapped, including joints, shears, and faults. Some variance between individuals' mapping efforts was noticed, and the effects of various magnetic influences upon a compass were examined. The examination of compass errors improved the credibility of the data. The compass analysis work is explained in Appendix A. Analysis of the fracture data will be presented in a future report

  6. Mineralogic and petrologic investigation of post-test core samples from the Spent Fuel Test - Climax

    International Nuclear Information System (INIS)

    Ryerson, F.J.; Beiriger, J.

    1985-02-01

    We have characterized a suite of samples taken subsequent to the end of the Spent Fuel Test - Climax by petrographic and microanalytical techniques and determined their mineral assemblage, modal properties, and mineral chemistry. The samples were obtained immediately adjacent to the canister borehole at a variety of depths and positions within the canister drift, as well as radially outward from each canister hole. This method of sampling allows variations in post-test mineralogic properties to be evaluated on the basis of (1) depth along a particular canister hole and (2) position within the canister drift, with respect to the heat and radiation sources, and with respect to the pre - test samples. In no case did we find any significant correlation between the mineralogical properties and variables listed above. In short, the Spent Fuel Test - Climax has produced no identifiable mineralogical response in the Climax quartz monzonite. 12 refs., 11 figs., 5 tabs

  7. Renormalizing Entanglement Distillation

    Science.gov (United States)

    Waeldchen, Stephan; Gertis, Janina; Campbell, Earl T.; Eisert, Jens

    2016-01-01

    Entanglement distillation refers to the task of transforming a collection of weakly entangled pairs into fewer highly entangled ones. It is a core ingredient in quantum repeater protocols, which are needed to transmit entanglement over arbitrary distances in order to realize quantum key distribution schemes. Usually, it is assumed that the initial entangled pairs are identically and independently distributed and are uncorrelated with each other, an assumption that might not be reasonable at all in any entanglement generation process involving memory channels. Here, we introduce a framework that captures entanglement distillation in the presence of natural correlations arising from memory channels. Conceptually, we bring together ideas from condensed-matter physics—ideas from renormalization and matrix-product states and operators—with those of local entanglement manipulation, Markov chain mixing, and quantum error correction. We identify meaningful parameter regions for which we prove convergence to maximally entangled states, arising as the fixed points of a matrix-product operator renormalization flow.

  8. Hoizontal retort for distilling

    Energy Technology Data Exchange (ETDEWEB)

    Archer, F; Papineau, W

    1854-12-15

    Improvements are disclosed in distilling peaty, schistose, bituminous and vegetable matters. These are arranging a retort in a horizontal position or so little inclined as not to permit matters charged at one end of the retort to fall readily to the other by gravity. The retort is heated externally by a fire at one end, near which end the spent products are discharged without opening the retort, which is done by the aid of two valves or slides, one being at all times closed when discharging products. The other end of the retort is provided with a divided hopper with two valves or slides, one of which is kept closed when the other is open, in order to charge the retort. Within is an endless chain carrying rakes so as to move the matters from the feeding to the discharging end. There are outlets for the distilled products furnished with condensers.

  9. Distillation apparatus for solid materials

    Energy Technology Data Exchange (ETDEWEB)

    Lurmann, F

    1860-06-26

    The distillation room is continuously charged by the charging mechanism with the material to be distilled. The distillation products pass into the chamber, where they are drawn out through the tube for purifying. The distillation residue is finally pushed out by the fresh material from the room and falls in the common room, from which it is removed through the air-tight door. In the canals enclosing the room heating gas circulates, which carries to the room the heat necessary for the distillation.

  10. Distillation, destructive: gas retorts

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, G; Buchanan, J

    1872-06-12

    Retorts used in the distillation of coal, cannel, shale, or other minerals are discharged by means of a joint metal bar or a chain inserted prior to or with the charge, and fitted with a plate or series of plates preferably with the back plate the largest. The rod or chain is formed with a hook, loop, or coupling for withdrawal, preferably by a steam windlass running on rails.

  11. Distilling carbonaceous materials

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, C A

    1924-04-15

    In apparatus of the kind set forth for distilling solid carbonaceous materials, a rotary retort in the form of a tubular, hollow cylindrical, or other similar hollow body, of small diameter, having a thin wall is provided to which the heat is applied externally, with means operative within it adapted, not only for cleaning the internal wall of the retort but also for distributing the heat throughout the mass of materials under treatment, substantially as described.

  12. Post test evaluation of a fire tested rail spent fuel cask

    International Nuclear Information System (INIS)

    Rack, H.J.; Yoshimura, H.R.

    1980-01-01

    Postmortem examination of a large rail-transported spent fuel shipping cask which had been exposed to a JP-4 fuel fire revealed the presence of two macrofissures in the outer cask shell. One, a part-through crack located within the seam weld fusion zone of the outer cask shell, is typical of hot cracks found in stainless steel weldments. The other, a through-crack, was apparently initiated during the formation of a copper-stainless steel dissimilar metal joint, with crack propagation through the cask outer shell having occurred during the fire-test. 8 figures

  13. Distilling solid carbonaceous materials

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, H; Laing, B

    1926-12-04

    In the distillation of solid carbonaceous materials with by-product recovery by direct heating with a gas such as water gas, producer gas, or combustion gas which is passed in counter-flow to the materials, the volume of the gas used is such as to lower the vapor tension of the volatiles to enable the oil vapor to be liberated at temperatures not exceeding 450 to 500/sup 0/C and so that the gaseous mixture may be cooled to from 80 to 100/sup 0/C without causing the highest boiling oil fraction to condense. Coking coals may be subjected to a preliminary heat treatment with gases containing an oxygen content of from 2 to 8 percent to reduce their coking properties, and oxygen may be added to the heating gases to assist the polymerization of resinous bodies. Lubricating oil may be obtained by treating the primary oil with caustic soda to remove tar acids, refining the residue with sulfuric acid, distilling off 25 percent of the refined oil and passing the remainder through a filter press at -5/sup 0/C to extract the paraffin wax. The residue of wax-free oil is distilled to yield a lubricating oil which at normal temperatures has a static coefficient of friction of from .1 to .185. Other specifications are referred to.

  14. Distillation of bituminous substances

    Energy Technology Data Exchange (ETDEWEB)

    du Buisson, M A.B.B.

    1845-06-23

    New and improved methods are described for the distillation of bituminous schistus and other bituminous substances, as well as for the purification, rectification, and preparation necessary for the employment of the productions obtained by such distillation for various useful purposes. This invention consists, first, in the arrangement and construction of furnace or apparatus for the distillation of schistus, and (any) other bituminous rocks. This furnace is made of circular brick-work, and is provided with the requisite number of fires placed round the circumference. The retort is of a conical or funnel shape, and when ready for use, has the appearance of one inverted cone being placed within a larger cone, in such a manner as to leave a space between the two cones for the reception of the schistus. Both cones are capable of being closed at their apexes, and their mouths, which are placed upwards, are connected together, and hermetically closed by a flat ring. The flame draft is caused to impinge against the lower portion of the outer cone and ascending, turns over the upper edge and descends within the inner cone to the mouth of the chimney, which is placed in the interior, and as low as may be convenient. A third cone is so placed within the inner cone as to cause the flame draft, in its descent, to bind against the surface of the retort.

  15. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  16. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  17. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  18. Power ramping test in the JMTR for PCI study of water reactor fuel

    International Nuclear Information System (INIS)

    Nakata, H.; Kanbara, M.; Ichikawa, M.

    1984-01-01

    Power ramping test is essential for PCI study of water reactor fuel. Boiling water capsules have been used for the tests in the JMTR. Heat generation of fuel rod in the capsule can be changed by the He-3 power control facility during reactor operation. Four specially designed fuel rods have been ramped to about 41-43 kW/m; two of them have small gaps filled with iodine, the other two are equipped with centerline temperature thermocouple. Fuel rod elongation detector is equipped to each capsule. For the fuel rods with small gap, unique contraction followed by ordinary fuel relaxation behaviour was observed right after the fast ramping. None of them failed. Future programme includes a series of tests of fuel rods irradiated in the high-pressure water loop at the JMTR and a verification test of remedy fuel which allows daily-load-following operation of BWRs. (author)

  19. Results of a diesel multiple unit fuel tank blunt impact test

    Science.gov (United States)

    2017-04-04

    The Federal Railroad Administrations Office of Research and Development is conducting research into passenger locomotive fuel tank crashworthiness. A series of impact tests is being conducted to measure fuel tank deformation under two types of dyn...

  20. Behavior of mixed-oxide fuel elements during the TOPI-1E transient overpower test

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.; Yamamoto, K.; Hirai, K.; Shikakura, S.

    1993-12-01

    A slow-ramp, extended overpower transient test was conducted on a group of nineteen preirradiated mixed-oxide fuel elements in EBR-II. During the transient two of the test elements with high-density fuel and tempered martensitic cladding (PNC-FMS) breached at an overpower of ∼75%. Fuel elements with austenitic claddings (D9, PNC316, and PNC150), many with aggressive design features and high burnups, survived the overpower transient and incurred little or no cladding strain. Fuel elements with annual fuel or heterogeneous fuel columns also behaved well

  1. Fission gas induced deformation model for FRAP-T6 and NSRR irradiated fuel test simulations

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hosoyamada, Ryuji; Mori, Yukihide

    1996-11-01

    Pulse irradiation tests of irradiated fuels under simulated reactivity initiated accidents (RIAs) have been carried out at the Nuclear Safety Research Reactor (NSRR). Larger cladding diameter increase was observed in the irradiated fuel tests than in the previous fresh fuel tests. A fission gas induced cladding deformation model was developed and installed in a fuel behavior analysis code, FRAP-T6. The irradiated fuel tests were analyzed with the model in combination with modified material properties and fuel cracking models. In Test JM-4, where the cladding temperature rose to higher temperatures and grain boundary separation by the pulse irradiation was significant, the fission gas model described the cladding deformation reasonably well. The fuel had relatively flat radial power distribution and the grain boundary gas from the whole radius was calculated to contribute to the deformation. On the other hand, the power density in the irradiated LWR fuel rods in the pulse irradiation tests was remarkably higher at the fuel periphery than the center. A fuel thermal expansion model, GAPCON, which took account of the effect of fuel cracking by the temperature profile, was found to reproduce well the LWR fuel behavior with the fission gas deformation model. This report present details of the models and their NSRR test simulations. (author)

  2. Test of Flow Characteristics in Tubular Fuel Assembly I - Establishment of test loop and measurement validation test

    International Nuclear Information System (INIS)

    Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

    2005-12-01

    Tubular type fuel has been developed as one of candidates for Advanced HANARO Reactor(AHR). It is necessary to test the flow characteristics such as velocity in each flow channels and pressure drop of tubular type fuel. A hydraulic test-loop to examine the hydraulic characteristics for a tubular type fuel has been designed and constructed. It consists of three parts; a) piping-loop including pump and motor, magnetic flow meter and valves etc, b) test-section part where a simulated tubular type fuel is located, and 3) data acquisition system to get reading signals from sensors or instruments. In this report, considerations during the design and installation of the facility and the selection of data acquisition sensors and instruments are described in detail. Before doing the experiment to measure the flow velocities in flow channels, a preliminary tests have been done for measuring the coolant velocities using pitot-tube and for validating the measurement accuracy as well. Local velocities of the radial direction in circular tubes are measured at regular intervals of 60 degrees by three pitot-tubes. Flow rate inside the circular flow channel can be obtained by integrating the velocity distribution in radial direction. The measured flow rate was compared to that of magnetic flow meter. According to the results, two values had a good agreement, which means that the measurement of coolant velocity by using pitot-tube and the flow rate measured by the magnetic flow meter are reliable. Uncertainty analysis showed that the error of velocity measurement by pitot-tube is less than ±2.21%. The hydraulic test-loop also can be adapted to others such as HANARO 18 and 36 fuel, in-pile system of FTL(Fuel Test Loop), etc

  3. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

    Energy Technology Data Exchange (ETDEWEB)

    Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

    2004-07-01

    We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

  4. Material gap membrane distillation: A new design for water vapor flux enhancement

    KAUST Repository

    Francis, Lijo; Ghaffour, NorEddine; Alsaadi, Ahmad Salem; Amy, Gary L.

    2013-01-01

    A new module design for membrane distillation, namely material gap membrane distillation (MGMD), for seawater desalination has been proposed and successfully tested. It has been observed that employing appropriate materials between the membrane

  5. Testing of low pressure proton exchange membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Bettoni, M; Naso, V; Lucentini, M; Rubini, L

    1998-07-01

    One of the main issues concerning PEMFC is the choice of operating pressure, for both stationary and automotive applications. This is because the air compressor may absorb a significant amount--up to 25%--of the power output of the fuel cells stack. A comparison has been made between the performance of various stacks of different dimensions, tested in the De Nora Laboratories operated at high (4 bar) and low (1.5 bar) pressures, considering power output reduced by the compressor power absorption. Differences of performance and efficiency between high and low pressure stacks have been noticed in the range of 10%. In operating at low pressure, higher efficiency is obtainable, but the maximum power of the stack is less; this means less fuel consumption, but requires a greater reacting surface and larger dimension of the stack. Consequently low pressures make the system simpler (a blower can be used instead of a compressor), and safer (there is practically no risk of breaking the membrane).

  6. Development, irradiation testing and PIE of UMo fuel at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.

    2005-01-01

    This paper reviews recent U-Mo dispersion fuel development, irradiation testing and postirradiation examination (PIE) activities at AECL. Low-enriched uranium fuel alloys and powders have been fabricated at Chalk River Labs, with compositions ranging from U-7Mo to U-10Mo. The bulk alloys and powders were characterized using optical and scanning electron microscopy, chemical analysis, X-ray diffraction and neutron diffraction analysis. The analyses confirmed that the powders were of high quality, and in the desired gamma phase. Subsequently, kilogram quantities of DU-Mo and LEU-Mo powder have been manufactured for commercial customers. Mini-elements have been fabricated with LEU-7Mo and LEU-10Mo dispersed in aluminum, with a nominal loading of 4.5 gU/cm 3 . These have been irradiated in the NRU reactor at linear powers up to 100 kW/m. The mini-elements achieved 60 atom% 235 U burnup in 2004 March, and the irradiation is continuing to a planned discharge burnup of 80 atom% 235 U. Interim PIE has been conducted on mini-elements that were removed after 20 atom% 235 U burnup. The PIE results are presented in this paper. (author)

  7. Vacuum distillation/vapor filtration water recovery

    Science.gov (United States)

    Honegger, R. J.; Neveril, R. B.; Remus, G. A.

    1974-01-01

    The development and evaluation of a vacuum distillation/vapor filtration (VD/VF) water recovery system are considered. As a functional model, the system converts urine and condensates waste water from six men to potable water on a steady-state basis. The system is designed for 180-day operating durations and for function on the ground, on zero-g aircraft, and in orbit. Preparatory tasks are summarized for conducting low gravity tests of a vacuum distillation/vapor filtration system for recovering water from urine.

  8. Performance testing of thermal analysis codes for nuclear fuel casks

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

  9. Test plan for Series 2 spent fuel cladding containment credit tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1984-10-01

    This test plan describes a second series of tests to be conducted by Westinghouse Hanford Company (WHC) to evaluate the effectiveness of breached cladding as a barrier to radionuclide release in the NNWSI-proposed geologic repository. These tests will be conducted at the Hanford Engineering Development Laboratory (HEDL). A first series of tests, initiated at HEDL during FY 1983, demonstrated specimen preparation and feasibility of the testing concept. The second series tests will be similar to the Series 1 tests with the following exceptions: NNWSI reference groundwater obtained from well J-13 will be used as the leachant instead of deionized water; fuel from a second source will be used; and certain refinements will be made in specimen preparation, sampling, and analytical procedures. 12 references, 5 figures, 5 tables

  10. The Hydraulic Test Procedure for Non-instrumented Irradiation Test Rig of Annular Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan; Park, Chan Kook

    2008-08-15

    This report presents the procedure of pressure drop test, vibration test and endurance test for the non-instrumented rig using the irradiation test in HANARO of advanced PWR annular fuel which were designed and fabricated by KAERI. From the out-pile thermal hydraulic tests, confirm the flow rate at the 200 kPa pressure drop and measure the RMS displacement at this time. And the endurance test is confirmed the wear and the integrity of the non-instrumented rig at the 110% design flow rate. This out-pile test perform the Flow-Induced Vibration and Pressure Drop Experimental Tester(FIVPET) facility. The instruments in FIVPET facility was calibrated in KAERI and the pump and the thermocouple were certified by manufacturer.

  11. CANFLEX fuel bundle cross-flow endurance test 2 (test procedure)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    This report describes test procedure of cross-flow 2 test for CANFLEX fuel. In October 1996. a cross-flow test was successfully performed in the KAERI Hot Test Loop for four hours at a water flow rate of 31kg/s, temperature of 266 deg C and inlet pressure of 11MPa, but it is requested more extended time periods to determine a realistic operational margin for the CANFLEX bundle during abnormal refuelling operations. The test shall be conducted for twenty two hours under the reactor conditions. After an initial period of ten hours, the test shall be stopped at the intervals of four hours for bundle inspection and inspect the test bundle end-plate to end-cap welds for failure or crack propagation using liquid penetrant examination. 2 refs., 1 fig. (Author)

  12. On the behaviour of dissolved fission gases prior to transient testing of fuel pins

    International Nuclear Information System (INIS)

    Wood, M.H.; Matthews, J.R.

    1978-10-01

    The TREAT and CABRI series of reactor safety experiments on irradiated fuel require the transfer of fuel pins from the reactor in which the fuel has achieved some burn-up to the test facility. Subsequently, the fuel is restored to power in the test facility for some time before transient heating is initiated. Such pre-test manoeuvres, where the fuel is subjected to changes in the fission rate and temperature, may have important consequences for the fission gas behaviour during the transient experiment. The results of rate theory calculations are used to assess these effects. (author)

  13. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  14. Hydrogen isotope separation by cryogenic distillation method

    International Nuclear Information System (INIS)

    Hayakawa, Nobuo; Mitsui, Jin

    1987-01-01

    Hydrogen isotope separation in fusion fuel cycle and tritium recovery from heavy water reactor are very important, and therefore the early establishment of these separation techniques are desired. The cryogenic distillation method in particular is promising for the separation of hydrogen isotope and the recovery of high concentrated tritium. The studies of hydrogen isotope separation by cryogenic distillation method have been carried out by using the experimental apparatus made for the first time in Japan. The separation of three components (H 2 -HD-D 2 ) under total reflux conditions was got by using the packing tower of 500 mm height. It was confirmed that the Height Equivalent Theoretical Plate (HETP) was 20 - 30 mm for the vapor's line velocity of 20 - 80 mm/s. (author)

  15. Determination of optimal wet ethanol composition as a fuel in spark ignition engine

    International Nuclear Information System (INIS)

    Fagundez, J.L.S.; Sari, R.L.; Mayer, F.D.; Martins, M.E.S.; Salau, N.P.G.

    2017-01-01

    Highlights: • Batch distillation to produce HEF and fuel blends of wet ethanol. • Conversion efficiency of a SI engine operating with HEF and wet ethanol. • NEF as a new metric to calculate the energy efficiency of HEF and wet ethanol. • Optimal wet ethanol composition as a fuel in SI engine based on NEF. - Abstract: Studies are unanimous that the greatest fraction of the energy necessary to produce hydrous ethanol fuel (HEF), i.e. above 95%v/v of ethanol in water, is spent on water removal (distillation). Previous works have assessed the energy efficiency of HEF; but few, if any, have done the same for wet ethanol fuel (sub-azeotropic hydrous ethanol). Hence, a new metric called net energy factor (NEF) is proposed to calculate the energy efficiency of wet ethanol and HEF. NEF calculates the ratio of Lower Heating Value (LHV) derived from ethanol fuel, total energy out, to energy used to obtain ethanol fuel as distillate, total energy in. Distillation tests were performed batchwise to obtain as distillate HEF and four different fuel blends of wet ethanol with a range from 60%v/v to 90%v/v of ethanol and the amount of energy spent to distillate each ethanol fuel calculated. The efficiency parameters of a SI engine operating with the produced ethanol fuels was tested to calculate their respective conversion efficiency. The results of net energy factors show a clear advantage of wet ethanol fuels over HEF; the optimal efficiency was wet ethanol fuel with 70%v/v of ethanol.

  16. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio programme

    International Nuclear Information System (INIS)

    Molecke, M.A.; Gregson, M.W.; Sorenson, K.B.

    2004-01-01

    We provide a detailed overview of an on-going, multinational test programme that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolised materials plus volatilised fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy/density device. The programme participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research programme. Sandia National Laboratories has the lead role for conducting this research programme; test programme support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. We provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests. We briefly summarise similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods. (author)

  17. PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

    International Nuclear Information System (INIS)

    Ganzmann, I.; Hille, D.; Staude, U.

    2009-01-01

    The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

  18. Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Julian F. [Thor Energy AS, Sommerrogaten 13-15, Oslo 0255 (Norway); Franceschini, Fausto [Westinghouse Electric Company LLC, 1000 Cranberry Woods Drive, Cranberry Township, PA 16066 (United States)

    2013-07-01

    Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cycle reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)

  19. Distilling coal, etc

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, W P

    1906-01-11

    Substances containing hydrocarbons, such as cannel coal, lignite, and shale, are destructively distilled by dividing the charge into small bodies confined in an air-tight chamber through which the products of combustion from a contiguous furnace are passed, the furnace serving also to heat the chamber. The temperature is kept below red heat so that the initial products, such as vapors of heavy oils, paraffin, waxes, naphthas, phenols, and cresols, are not decomposed and there is no formation of gaseous products such as naphthalene and benzol. The operation is of short duration, and large amounts of good coke are produced.

  20. Detail design of test loop for FIV in fuel bundle and preliminary test

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gunl; Lee, Wan Young; Kim, Sung Won [Hannam University, Taejeon (Korea)

    2002-04-01

    It is urgent to develop the analytical model for the structural/mechanical integrity of fuel rod. In general, it is not easy to develop a pure analytical model. Occasionally, experimental results have been utilized for the model.Because of this reason, it is required to design proper test loop. Using the optimized test loop, With the optimized test loop, the dynamic behaviour of the rod will be evaluated and the critical flow velocity, which the rod loses the stability in, will be measured for the design of the rod. To verify the integrity of the fuel rod, it is required to evaluate the dynamic behaviour and the critical flow velocity with the test loop. The test results will be utilized to the design of the rod. Generally, the rod has a ground vibration due to turbulence in wide range of flow velocity and the amplitude of vibration becomes larger by the resonance, in a range of the velocity where occurs vortex. The rod loses stability in critical flow velocity caused by fluid-elastic instability. For the purpose of the present work to perform the conceptional design of the test loop, it is necessary (1) to understand the mechanism of the flow-induced vibration and the related experimental coefficients, (2) to evaluate the existing test loops for improving the loop with design parameters and (3) to decide the design specifications of the major equipments of the loop. 35 refs., 14 figs., 4 tabs. (Author)

  1. Fuel dynamics loss-of-flow test L3. Final report

    International Nuclear Information System (INIS)

    Fischer, A.K.; Lo, R.K.; Barts, E.W.

    1976-06-01

    The behavior of FTR-type, mixed-oxide, preirradiated, ''intermediate-power-structure'' fuel during a simulation of an FTR loss-of-flow accident was studied in the Mark-IIA integral TREAT loop. Analysis of the data reported here leads to a postulated scenario (sequence and timing) of events in the test. This scenario is presented, together with the calculated timing of events obtained by use of the SAS code. The initial fuel motion, starting during the preheat phase, consisted of coherent motion of the entire intact fuel bundle toward the pump. Incoherence developed as temperature rose. The fuel motion was mostly upward, and the greatest was in the top third of the fuel column. Fuel fragments formed against the pump side of the fluted tube near the original fuel midplane. A penetration of fluted tube occurred. A sudden voiding of the central region of the fuel column occurred at 29.75 s and was largely completed within 150 ms. The lower steel blockage of the fuel elements occurred in the vicinity of the lower insulator pellets. The upper steel blockage just above the tops of the original fuel pins appeared to have channels through it. Cladding and spacer wires melted away in the fuel section. Fuel pellets were only evident at and above the top and at the bottom of the original fuel column, where a large mass of melted fuel was present. Over the length of the fuel column, most of the fluted tube had melted away

  2. Estimation of metallic impurities in uranium by carrier distillation method

    International Nuclear Information System (INIS)

    Page, A.G.; Godbole, S.V.; Deshkar, S.B.; Joshi, B.D.

    1976-01-01

    An emission spectrographic method has been standardised for the estimation of twenty-two metallic impurities in uranium using carrier-distillation technique. Silver chloride with a concentration of 5% has been used as the carrier and palladium and gallium are used as internal standards. Precision and accuracy determinations of the synthetic samples indicate 6-15% deviation for most of the elements. Using the method described here, five uranium reference samples received from C.E.A.-France were analysed. The detection limits obtained for Cd, Co and W are lower than those reported in the literature while limits for the remaining elements are comparable to the values reported. The method is suitable for the chemical quality control analysis of uranium used for the Fast Breeder Test Reactor (FBTR) fuel. (author)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  4. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  5. Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

    International Nuclear Information System (INIS)

    Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

    1992-01-01

    The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

  6. Test experiences with the DaimlerChrysler: Fuel cell electric vehicle NECAR

    OpenAIRE

    Friedlmeier Gerardo; Friedrich J.; Panik F.

    2002-01-01

    The DalmlerChrysler fuel cell electric vehicle NECAR 4, a hydrogen-fueled zero-emission compact car based on the A-Class of Mercedes-Benz, is described. Test results obtained on the road and on the dynamometer are presented. These and other results show the high technological maturity reliability and durability already achieved with fuel cell technology.

  7. Test experiences with the DaimlerChrysler: Fuel cell electric vehicle NECAR

    Directory of Open Access Journals (Sweden)

    Friedlmeier Gerardo

    2002-01-01

    Full Text Available The DalmlerChrysler fuel cell electric vehicle NECAR 4, a hydrogen-fueled zero-emission compact car based on the A-Class of Mercedes-Benz, is described. Test results obtained on the road and on the dynamometer are presented. These and other results show the high technological maturity reliability and durability already achieved with fuel cell technology.

  8. Failure analysis of burst tested fuel tube samples

    International Nuclear Information System (INIS)

    Padmaprabu, C.; Ramana Rao, S.V.; Srivatsava, R.K.

    2005-01-01

    The Total Circumferential Elongation (TCE) is an important parameter for evaluation of ductility of the Zircaloy-4 fuel tubes for the PHWR reactors. The TCE values of the fuel tubes were obtained using the burst testing technique. In some lots there is a variation in the values of the TCE. To investigate the reasons for such a large variation in the TCE, samples were selected at appropriate intervals and sectioned at the fractured portion. The surface morphology of the fractured surfaces was examined under Scanning Electron Microscope (SEM) equipped with Energy Dispersive Spectrometer (EDS). The morphologies show segregation of elements at specific locations. Energy dispersive spectra was obtained from those segregated particles. According to the magnitude of TCE value the samples were classified into low, intermediate and high ductility. Low ductility samples were found to contain large amount of segregations along the thickness direction of the tube. This forms a brittle region and a path for the easy crack growth along thickness direction. In the case of intermediate samples the segregation occurred in fewer locations compared to low ductile samples and also confined to the circumferential direction of the outside surface of the tube. Due to this, probability of crack formation at the surface of the tube could be high. But crack growth would be slower in the ductile matrix along the thickness direction resulting in the enhancement of TCE value compared to the low ductile sample. In the high ductile samples, the segregations were very scarce and found to be isolated and embedded in the ductile matrix. The mode of failure in these types of samples was found to be purely ductile. Cracks were found to originate solely from the micro voids in the material. As the probability of crack formation and its propagation is low, very high TCE values were observed in these samples. Microstructural observations of fractured surfaces and EDAX analysis was able to identify the

  9. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  10. Change in geometrical parameters of WWER high burnup fuel rods under operational conditions and transient testing

    International Nuclear Information System (INIS)

    Kanashov, B.; Amosov, S.; Lyadov, G.; Markov, D.; Ovchinnikov, V; Polenok, V.; Smirnov, A.; Sukhikh, A.; Bek, E.; Yenin, A.; Novikov, V.

    2001-01-01

    The paper discusses changes in fuel rods geometric parameters as result of operation conditions and burnups. The degree of geometry variability of fuel rods, cladding and column is one of the most important characteristics affecting fuel serviceability. On the other hand, changes in fuel rod geometric parameters influence fuel temperature, fission gas release, fuel-to-cladding stress strained state as well as the degree of interaction with FA skeleton elements and skeleton rigidity. Change in fuel-to-cladding gap is measured using compression technique. The axial distribution of fuel-to-cladding gap demonstrates the largest decrease of the gap in the region 500 to 2000 mm from the bottom of the fuel rod (WWER-440) and in the region of 500 to 3000 mm for WWER-1000. The cladding material creep in WWER fuel rods together with the radiation growth results in fuel rod cladding elongation. A set of transient tests for spent WWER-440 and WWER-1000 fuel rods carried out in SSC RIAR during a period 1995-1999, with the aim to estimate the changes in geometric parameters of FRs. The estimation of changes in outer diameter of cladding and fuel column and fuel-to-cladding gap are performed in transient conditions (changes in linear power range of 180 to 400 W/cm) for both WWER-440 and WWER-1000. WWER-440 fuel rods having the same burnup and close fuel-cladding contact before testing are subjected to considerable hoop cladding strain in testing up to 300 W/cm. But the hoop strain does not grow due to the structural changes in fuel column and decrease in central hole diameter occurred when the power is higher

  11. Distillation with Sublogarithmic Overhead

    Science.gov (United States)

    Hastings, Matthew B.; Haah, Jeongwan

    2018-02-01

    It has been conjectured that, for any distillation protocol for magic states for the T gate, the number of noisy input magic states required per output magic state at output error rate ɛ is Ω [log (1 /ɛ )] . We show that this conjecture is false. We find a family of quantum error correcting codes of parameters ⟦ ∑ i =w +1 m (m i ),∑ i =0 w (m i ),∑ i =w +1 r +1 (r +1 i )⟧ for any integers m >2 r , r >w ≥0 , by puncturing quantum Reed-Muller codes. When m >ν r , our code admits a transversal logical gate at the ν th level of Clifford hierarchy. In a distillation protocol for magic states at the level ν =3 (T gate), the ratio of input to output magic states is O (logγ(1 /ɛ )) , where γ =log (n /k )/log (d )<0.678 for some m , r , w . The smallest code in our family for which γ <1 is on ≈258 qubits.

  12. Distilling carbonaceous materials

    Energy Technology Data Exchange (ETDEWEB)

    Garrow, J R

    1921-04-16

    To obtain an increased yield of by-products such as oils, ammonia, and gas from coal, oil shale, wood, peat, and the like by low and medium temperature processes, the requisite quantity of hot producer gas from a gas producer, is caused to travel, without ignition, through the material as it passes in a continuous manner through the retort so that the sensible heat of the producer gas is utilized to produce distillation of the carbonaceous material, the gases passing to a condenser, absorption apparatus, and an ammonia absorber respectively. In a two-stage method of treatment of materials such as peat or the like, separate supplies of producer gas are utilized for a preliminary drying operation and for the distillation of the material, the drying receptacle and the retort being joined together to render the process continuous. The gas from the drying receptacle may be mixed with the combined producer and retort gas from the retort, after the hydrocarbon oils have deen removed therefrom.

  13. Distilling carbonaceous materials

    Energy Technology Data Exchange (ETDEWEB)

    Ironside, T G

    1921-09-01

    In the distillation of carbonaceous material such as shale, coal, lignite, wood or liquid hydrocarbons, the material is mixed with a heated granular substance such as sand which supplies the necessary heat. The shale or the like, which may be preheated, is fed from a hopper by a worm conveyer to a tube leading into a retort, and the heated granular material such as sand is supplied from a jacketed container through a tube. On the lower end of a rotary shaft are radial arms to which are fixed angularly disposed blades which serve to mix the shale and hot sand and deliver the residue to a central discharge pipe closed at the bottom by a conical valve which opens when the weight of the superimposed material is sufficient. The distillates are taken off by an outlet. Steam vapor or gas may be supplied to the retort, preferably through a hollow shaft leading to hollow stirrers perforated to permit of the gas passing into the material. The retort may be externally heated by hot gases in the space surrounding the retort, and the latter may be divided by horizontal floors so that the material is caused to funnel from the periphery to the center of the floor, then through a central opening on to the floor next below, and from the center to the periphery of this floor, and so on.

  14. Retorts; distillation, destructive

    Energy Technology Data Exchange (ETDEWEB)

    Kirk, A C

    1869-11-09

    Relates to retorts for distilling shale or other purposes. Each retort has a mouthpiece, which projects into a water trough, having a curved bottom, and sides rising up to ridges, from which extend shoots. Attached to links is the reciprocating discharge bar, actuated from a steam engine, etc., which, through worm-wheel gearing and a shaft, imparts an intermittent motion to a pawl which propels a ratchet-wheel, which is prevented from turning backward by a catch, and drives a rotary shaft, the motion being transmitted to a bar through a crank and rod. The bar, in rising, pushes the residue from the retort out of the trough, over the ridges, and down the shoots into the truck, etc. The vertical retort is made in two lengths, at the joint between which is held the flange of the shell, joined to, or cast with, the part of the retort and forming a jacket into which steam or other vapor or gas is introduced by a pipe. The steam, etc., circulates through an external spiral rib of the shell and issues from the bottom of the shale among the shale, etc., being distilled.

  15. Qutrit magic state distillation

    International Nuclear Information System (INIS)

    Anwar, Hussain; Browne, Dan E; Campbell, Earl T

    2012-01-01

    Magic state distillation (MSD) is a purification protocol that plays an important role in fault-tolerant quantum computation. Repeated iteration of the steps of an MSD protocol generates pure single non-stabilizer states, or magic states, from multiple copies of a mixed resource state using stabilizer operations only. Thus mixed resource states promote the stabilizer operations to full universality. MSD was introduced for qubit-based quantum computation, but little has been known concerning MSD in higher-dimensional qudit-based computation. Here, we describe a general approach for studying MSD in higher dimensions. We use it to investigate the features of a qutrit MSD protocol based on the five-qutrit stabilizer code. We show that this protocol distils non-stabilizer magic states, and identify two types of states that are attractors of this iteration map. Finally, we show how these states may be converted, via stabilizer circuits alone, into a state suitable for state-injected implementation of a non-Clifford phase gate, enabling non-Clifford unitary computation. (paper)

  16. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

  17. 40 CFR 86.335-79 - Gasoline-fueled engine test cycle.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Gasoline-fueled engine test cycle. 86....335-79 Gasoline-fueled engine test cycle. (a) The following test sequence shall be followed in... operating the engine at the higher approved load setting during cycle 1 and at the lower approved load...

  18. Ground test facilities for evaluating nuclear thermal propulsion engines and fuel elements

    International Nuclear Information System (INIS)

    Allen, G.C.; Beck, D.F.; Harmon, C.D.; Shipers, L.R.

    1992-01-01

    Interagency panels evaluating nuclear thermal propulsion development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and design issues of a proposed ground test complex for evaluating nuclear thermal propulsion engines and fuel elements being developed for the Space Nuclear Thermal Propulsion (SNTP) program. 2 refs

  19. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

  20. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    International Nuclear Information System (INIS)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh

    1995-07-01

    This is the '94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author)