WorldWideScience

Sample records for fuels spill test

  1. 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

  2. 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

  3. An evaluation of propane as a fuel for testing fire-resistant oil spill containment booms

    International Nuclear Information System (INIS)

    Walton, W. D.; Twilley, W. H.

    1997-01-01

    A series of experiments have been conducted to measure and compare the thermal exposure to a fire-resistant boom from liquid hydrocarbon fuel and propane fires. The objective was to test the potential of propane fueled fires as a fire source for testing fire-resistant oil spill containment booms.Thermal exposure from propane fires have been measured with and without waves. Results indicated that although propane diffusion flames on water look like liquid hydrocarbon fuel flames and produce very little smoke, the heat flux at the boom location from propane fires is about 60 per cent of that from liquid hydrocarbon fuel fires. Despite the attractive features in terms of ease of application, control and smoke emissions, it was concluded that the low heat flux would preclude the application of propane as a fuel for evaluating fire resistant containment booms. 2 refs., 7 figs

  4. Responding effectively to fuel spills at airports

    International Nuclear Information System (INIS)

    Williams, L.E.

    1991-01-01

    Fuel spills are among the most frequent causes of emergency calls faced by airport firefighters. Most fuel spills are a result of human error and careless procedures. They always constitute an emergency and require fast, efficient action to prevent disaster. A fuel spill is an accidental release of fuel, in this case, from an aircraft fuel system, refueling vehicle or refueling system. A normal release of a few drops of fuel associated with a disconnection or other regular fueling operations should not be classified as a fuel spill. However, anytime fuel must be cleaned up and removed from an area, a fuel spill has occurred. Volatile fuels pose significant threats to people, equipment, facilities and cargo when they are released. Anyone near a spill, including ramp workers, fueling personnel and aircraft occupants, are in danger if the fuel ignites. Buildings and equipment in a spill area, such as terminals, hangars, aircraft, fuel trucks and service equipment also are at risk. An often neglected point is that aircraft cargo also is threatened by fuel spills

  5. Cleanup of a jet fuel spill

    Science.gov (United States)

    Fesko, Steve

    1996-11-01

    Eaton operates a corporate aircraft hanger facility in Battle Creek, Michigan. Tests showed that two underground storage tanks leaked. Investigation confirmed this release discharged several hundred gallons of Jet A kerosene into the soil and groundwater. The oil moved downward approximately 30 feet and spread laterally onto the water table. Test results showed kerosene in the adsorbed, free and dissolved states. Eaton researched and investigated three clean-up options. They included pump and treat, dig and haul and bioremediation. Jet fuel is composed of readily biodegradable hydrocarbon chains. This fact coupled with the depth to groundwater and geologic setting made bioremediation the low cost and most effective alternative. A recovery well was installed at the leading edge of the dissolved contamination. A pump moved water from this well into a nutrient addition system. Nutrients added included nitrogen, phosphorous and potassium. Additionally, air was sparged into the water. The water was discharged into an infiltration gallery installed when the underground storage tanks were removed. Water circulated between the pump and the infiltration basin in a closed loop fashion. This oxygenated, nutrient rich water actively and aggressively treated the soils between the bottom of the gallery and the top of the groundwater and the groundwater. The system began operating in August of 1993 and reduced jet fuel to below detection levels. In August of 1995 The State of Michigan issued a clean closure declaration to the site.

  6. Effectiveness testing of spill-treating agents

    International Nuclear Information System (INIS)

    Fingas, M.F.; Stoodley, R.; Laroche, N.

    1990-01-01

    Laboratory effectiveness tests are described for four classes of spill-treating agents: solidifiers, demulsifying agents, surface-washing agents and dispersants. Many treating agents in these four categories have been tested for effectiveness and the results are presented. Solidifiers or gelling agents solidify oil, requiring a large amount of agent to solidify oil-ranging between 16% by weight, to over 200%. Emulsion breakers prevent or reverse the formation of water-in-oil emulsions. A newly-developed effectiveness test shows that only one product is highly effective; however, many products will work, but require large amounts of spill-treating agent. Surfactant--containing materials are of two types, surface-washing agents and dispersants. Testing has shown that an agent that is a good dispersant is conversely a poor surface-washing agent, and vice versa. Tests of surface-washing agents show that only a few agents have effectiveness of 25-40%, where this effectiveness is the percentage of heavy oil removed from a test surface. Results using the 'swirling flask' test for dispersant effectiveness are reported. Heavy oils show effectiveness values of about 1%, medium crudes of about 10%, light crude oils of about 30% and very light oils of about 90%. (author)

  7. Validation of an orimulsion spill fates model using observations from field test spills

    International Nuclear Information System (INIS)

    French, D. P.; Rines, H.; Masciangioli, P.

    1997-01-01

    The SIMAP Spill Impact Model system was developed to simulate fates and effects of spilled oil and other fuels in 3-D and time. Orimulsion is a Venezuelan product consisting of 70 per cent bitumen and 30 per cent water which has been shipped to many parts of the world for some time without an accidental spill into coastal or marine waters. In July 1966 two intentional spills of Orimulsion into Carribean waters were made and sampled in detail in order to verify the SIMAP model. Data on physical dispersion was collected at the same time. Data collected in the field was compared with model simulations. Results confirmed SIMAP's ability to predict the increasing dispersion and shearing of the bitumen plume as wind speed increases, as well as the actual field distribution of subsurface and surface bitumen. 17 refs., 7 tabs., 26 figs

  8. Bioremediation Potential of Terrestrial Fuel Spills

    OpenAIRE

    Song, Hong-Gyu; Wang, Xiaoping; Bartha, Richard

    1990-01-01

    A bioremediation treatment that consisted of liming, fertilization, and tilling was evaluated on the laboratory scale for its effectiveness in cleaning up a sand, a loam, and a clay loam contaminated at 50 to 135 mg g of soil−1 by gasoline, jet fuel, heating oil, diesel oil, or bunker C. Experimental variables included incubation temperatures of 17, 27, and 37°C; no treatment; bioremediation treatment; and poisoned evaporation controls. Hydrocarbon residues were determined by quantitative gas...

  9. Fuel spill identification by gas chromatography -- genetic algorithms/pattern recognition techniques

    International Nuclear Information System (INIS)

    Lavine, B.K.; Moores, A.J.; Faruque, A.

    1998-01-01

    Gas chromatography and pattern recognition methods were used to develop a potential method for typing jet fuels so a spill sample in the environment can be traced to its source. The test data consisted of 256 gas chromatograms of neat jet fuels. 31 fuels that have undergone weathering in a subsurface environment were correctly identified by type using discriminants developed from the gas chromatograms of the neat jet fuels. Coalescing poorly resolved peaks, which occurred during preprocessing, diminished the resolution and hence information content of the GC profiles. Nevertheless a genetic algorithm was able to extract enough information from these profiles to correctly classify the chromatograms of weathered fuels. This suggests that cheaper and simpler GC instruments ca be used to type jet fuels

  10. Bioremediation of a No. 6 fuel spill

    International Nuclear Information System (INIS)

    Fogel, S.; Leahy, M.; Jones, M.; Butts, R.

    1991-01-01

    Although it is widely recognized that the major constituents of petroleum are highly biodegradable, the natural or unenhanced rate can be extremely slow. This is best exemplified by the petroleum reserves which have existed for million of years without substantial biodegradation due exclusively to nutrient limitations. The limiting nutrients include oxygen, nitrogen, phosphorus, and trace elements. The enhancement of the biodegradation process is termed bioremediation and consists of adding these nutrients in a prescribed and defined manner to soil and aquifers. Laboratory biodegradation tests are conducted prior to pilot- or full-scale remedial action to ensure the feasibility of the process. Depending on the comparability between the laboratory test and the field application, the data generated from the laboratory scale test can be used for purposes of field design and for prediction of the rate of biodegradation under field conditions. It is a critical assumption in the remediation industry that a laboratory treatment simulation does indeed simulate the field process and predicts the results of the full-scale remediation. This paper provides evidence that a laboratory scale treatment simulation can indeed predict field results

  11. Oil spill sorbents: Testing protocol and certification listing program

    International Nuclear Information System (INIS)

    Cooper, D.; Gausemel, I.

    1993-01-01

    Environment Canada's Emergencies Engineering Division is spearheading a program in conjunction with the Canadian General Standards Board that would see the development of a certification and listing program in addition to a national standard for the testing of sorbent materials. Funding for this program is provided by Environment Canada (EC), Canadian Coast Guard (CCG), Marine Spill Response Corporation (MSRC), US Coast Guard (USCG), and US Minerals Management Service (MMS). The test methods are based upon those defined by the American Society for Testing and Materials and previous test methods developed by Environment Canada for our series of reports entitled Selection Criteria and Laboratory Evaluation of Oil Spill Sorbents. This series, which was started in 1975, encompasses a number of commercially available oil spill sorbents tested with different petroleum products and hydrocarbon solvents. The testing program will categorize the sorbents according to their operating characteristics. The main categories are oil spills on water, oil spills on land, and industrial use. The characteristics to be evaluated with the new test protocols include initial and maximum sorption capacities, water pickup, buoyancy, reuse potential, retention profile, disintegration (material integrity), and ease of application and retrieval. In the near future are plans to incorporate changes to the test that would involve increasing the list of test liquids to encompass spills in an industrial setting, in addition to testing sorbent booms and addressing the disposal problem

  12. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 398: Area 25 Spill Sites, Nevada Test Site, Nevada; TOPICAL

    International Nuclear Information System (INIS)

    K. B. Campbell

    2001-01-01

    This Streamlined Approach for Environmental Restoration (SAFER) plan addresses the activities necessary to close Corrective Action Unit (CAU) 398: Area 25 Spill Sites. CAU 398, located in Area 25 of the Nevada Test Site, is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) (FFACO, 1996), and consists of the following 13 Corrective Action Sites (CASs) (Figure 1): (1) CAS 25-44-01 , a fuel spill on soil that covers a concrete pad. The origins and use of the spill material are unknown, but the spill is suspected to be railroad bedding material. (2) CAS 25-44-02, a spill of liquid to the soil from leaking drums. (3) CAS 25-44-03, a spill of oil from two leaking drums onto a concrete pad and surrounding soil. (4) CAS 25-44-04, a spill from two tanks containing sulfuric acid and sodium hydroxide used for a water demineralization process. (5) CAS 25-25-02, a fuel or oil spill from leaking drums that were removed in 1992. (6) CAS 25-25-03, an oil spill adjacent to a tipped-over drum. The source of the drum is not listed, although it is noted that the drum was removed in 1991. (7) CAS 25-25-04, an area on the north side of the Engine-Maintenance, Assembly, and Disassembly (E-MAD) facility, where oils and cooling fluids from metal machining operations were poured directly onto the ground. (8) CAS 25-25-05, an area of oil and/or hydraulic fluid spills beneath the heavy equipment once stored there. (9) CAS 25-25-06, an area of diesel fuel staining beneath two generators that have since been removed. (10) CAS 25-25-07, an area of hydraulic oil spills associated with a tunnel-boring machine abandoned inside X-Tunnel. (11) CAS 25-25-08, an area of hydraulic fluid spills associated with a tunnel-boring machine abandoned inside Y-Tunnel. (12) CAS 25-25-16, a diesel fuel spill from an above-ground storage tank located near Building 3320 at Engine Test Stand-1 (ETS-1) that was removed in 1998. (13) CAS 25-25-17, a hydraulic oil spill

  13. Environmental Assessment for the LGF Spill Test Facility at Frenchman Flat, Nevada Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Patton, S.E.; Novo, M.G.; Shinn, J.H.

    1986-04-01

    The LGF Spill Test Facility at Frenchman Flat, Nevada Test Site, is being constructed by the United States Department of Energy (DOE). In this Environmental Assessment, environmental consequences of spilling hazardous materials in the Frenchman Flat basin are evaluated and mitigations and recommendations are stated in order to protect natural resources and reduce land-use impacts. Guidelines and restrictions concerning spill-test procedures will be determined by the LGF Test Facility Operations Manager and DOE based on toxicity documentation for the test material, provided by the user, and mitigations imposed by the Environmental Assessment. In addition to Spill Test Facility operational procedures, certain assumptions have been made in preparation of this document: no materials will be considered for testing that have cumulative, long-term persistence in the environment; spill tests will consist of releases of 15 min or less; and sufficient time will be allowed between tests for recovery of natural resources. Geographic limits to downwind concentrations of spill materials were primarily determined from meteorological data, human occupational exposure standards to hazardous materials and previous spill tests. These limits were established using maximum spill scenarios and environmental impacts are discussed as worst case scenarios; however, spill-test series will begin with smaller spills, gradually increasing in size after the impacts of the initial tests have been evaluated.

  14. Environmental Assessment for the LGF Spill Test Facility at Frenchman Flat, Nevada Test Site

    International Nuclear Information System (INIS)

    Patton, S.E.; Novo, M.G.; Shinn, J.H.

    1986-04-01

    The LGF Spill Test Facility at Frenchman Flat, Nevada Test Site, is being constructed by the United States Department of Energy (DOE). In this Environmental Assessment, environmental consequences of spilling hazardous materials in the Frenchman Flat basin are evaluated and mitigations and recommendations are stated in order to protect natural resources and reduce land-use impacts. Guidelines and restrictions concerning spill-test procedures will be determined by the LGF Test Facility Operations Manager and DOE based on toxicity documentation for the test material, provided by the user, and mitigations imposed by the Environmental Assessment. In addition to Spill Test Facility operational procedures, certain assumptions have been made in preparation of this document: no materials will be considered for testing that have cumulative, long-term persistence in the environment; spill tests will consist of releases of 15 min or less; and sufficient time will be allowed between tests for recovery of natural resources. Geographic limits to downwind concentrations of spill materials were primarily determined from meteorological data, human occupational exposure standards to hazardous materials and previous spill tests. These limits were established using maximum spill scenarios and environmental impacts are discussed as worst case scenarios; however, spill-test series will begin with smaller spills, gradually increasing in size after the impacts of the initial tests have been evaluated

  15. Fate of dispersed marine fuel oil in sediment under pre-spill application strategy

    International Nuclear Information System (INIS)

    Jian Hua

    2004-01-01

    A comparison of the movement of dispersed oil in marine sediment under two dispersant application scenarios, applied prior to and after oil being spilled overboard, was examined. The pre-spill application scenario caused much less oil to be retained in the top sediment than post-spill scenario. The difference in oil retention in the top sediment between pre- and post-spill application scenario increased with increase in fuel oil temperature. For fuel oil above 40 o C, the difference in the effect of pre-spill application strategy under various water temperatures was negligible. When soap water was used as replacement for chemical dispersant, almost one-half as much oil was retained in the top sediment as that when using chemical dispersant. The adsorption of dispersed oil to the top sediment was almost proportionally decreased with doubling of soap dosage. (Author)

  16. Laboratory tests, experimental oil spills, models, and reality: The Braer oil spill

    International Nuclear Information System (INIS)

    Reed, M.; Daling, P.S.; Brandvik, P.J.; Singsaas, I.

    1993-01-01

    The IKU Petroleum Research organization in Norway has accumulated data on the weathering behavior of spilled oils and petroleum products, mainly pertaining to North Sea crudes. Recent weathering research at IKU has been carried out in an elliptical mesoscale flume and in field tests consisting of experimental releases of crude oil. Results of these tests provided information on oil spill dispersion, evaporation, and emulsification. When the tanker Braer grounded in the Shetland Islands in January 1993 in extreme environmental conditions, the imminent release of a load of 84,000 tonnes of North Sea oil confronted response personnel with a variety of issues including the use of dispersants as a response action. Relevant information on the expected behavior of the crude was obtained within a day of the grounding as a result of close relations between IKU and Warren Spring Laboratory. The question is raised whether such information, which could have been spread between several organizations around the world, will be rapidly accessible in the event of another major spill. It is proposed to establish an electronically accessible database on the behavior and fate of specific oils and petroleum products to address this problem. 9 refs., 4 figs

  17. Predicting the weathering of fuel and oil spills: A diffusion-limited evaporation model.

    Science.gov (United States)

    Kotzakoulakis, Konstantinos; George, Simon C

    2018-01-01

    The majority of the evaporation models currently available in the literature for the prediction of oil spill weathering do not take into account diffusion-limited mass transport and the formation of a concentration gradient in the oil phase. The altered surface concentration of the spill caused by diffusion-limited transport leads to a slower evaporation rate compared to the predictions of diffusion-agnostic evaporation models. The model presented in this study incorporates a diffusive layer in the oil phase and predicts the diffusion-limited evaporation rate. The information required is the composition of the fluid from gas chromatography or alternatively the distillation data. If the density or a single viscosity measurement is available the accuracy of the predictions is higher. Environmental conditions such as water temperature, air pressure and wind velocity are taken into account. The model was tested with synthetic mixtures, petroleum fuels and crude oils with initial viscosities ranging from 2 to 13,000 cSt. The tested temperatures varied from 0 °C to 23.4 °C and wind velocities from 0.3 to 3.8 m/s. The average absolute deviation (AAD) of the diffusion-limited model ranged between 1.62% and 24.87%. In comparison, the AAD of a diffusion-agnostic model ranged between 2.34% and 136.62% against the same tested fluids. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. NAPL migration and ecotoxicity of conventional and renewable fuels in accidental spill scenarios.

    Science.gov (United States)

    Malk, Vuokko; Barreto Tejera, Eduardo; Simpanen, Suvi; Dahl, Mari; Mäkelä, Riikka; Häkkinen, Jani; Kiiski, Anna; Penttinen, Olli-Pekka

    2014-01-01

    Fuels derived from non-petroleum renewable resources have raised interest due to their potential in replacing petroleum-based fuels, but information on their fate and effects in the terrestrial and aquatic environments in accidental spill scenario is limited. In this study, migration of four fuels (conventional diesel, conventional gasoline, renewable diesel NExBTL, and ethanol-blended gasoline RE85 containing maximum 85% ethanol) as non-aqueous phase liquids (NAPL) in soil was demonstrated in a laboratory-scale experiment. Ecotoxicity data was produced for the same fuels. There was no significant difference in migration of conventional and renewable diesel, but gasoline migrated 1.5 times deeper and 7-9 times faster in sand than diesel. RE85 spread horizontally wider but not as deep (p gasoline. Conventional gasoline was the most toxic (lethal concentration [LC50] 20 mg/kg total hydrocarbon content [THC]) among the studied fuels in soil toxicity test with earthworm Eisenia fetida followed by ethanol-blended gasoline (LC50 1,643 mg/kg THC) and conventional diesel (LC50 2,432 mg/kg THC), although gasoline evaporated fast from soil. For comparison, the toxicity of the water-accommodated fractions (WAF) of the fuels was tested with water flea Daphnia magna and Vibrio fischeri, also demonstrating groundwater toxicity. The WAF of conventional gasoline and RE85 showed almost similar toxicity to both the aquatic test species. EC50 values of 1:10 (by volume) WAF were 9.9 %WAF (gasoline) and 9.3 %WAF (RE85) to D. magna and 9.3 %WAF (gasoline) and 12.3 %WAF (RE85) to V. fischeri. Low solubility decreased toxicity potential of conventional diesel in aquatic environment, but direct physical effects of oil phase pose a threat to organisms in nature. Renewable diesel NExBTL did not show clear toxicity to any test species.

  19. Domestic fuel oil spill prevention committee : report to Ministers of Government Services and Lands and Environment

    International Nuclear Information System (INIS)

    2001-07-01

    The number of reported spills from domestic fuel oil systems increased dramatically in Newfoundland, which prompted the Minister of Government Services and Lands to arrange a meeting with representatives from consumers, the fuel service industry and the insurance industry to ensure proper measures were taken for the prevention of domestic fuel spills. A joint committee consisting of industry and government representatives was formed as a result of this meeting, to examine and investigate the situation and report to the Minister of Government Services and Lands. Advice on means to address the problems associated with domestic fuel oil spills was provided, as well as mechanisms to minimize such occurrences in the future. Also included in the review were small commercial storage tanks units of no more than 2500 litres, as small commercial establishments often have heating systems similar in size to residential units. Gradual leaks that go undetected for years often occur, as do the catastrophic rupture of the fuel storage tank itself. Rusting and exterior tubing are some of the causes of spills. The contamination of surrounding soil and/or groundwater can occur as a result of the spills, and fumes can enter residences through foundation walls of the sewer system. Condensation within the tank can lead to corrosion of the fuel tanks. A number of recommendations were made in the report, such as the establishment of regulations pertaining to the construction, installation, servicing and fueling of domestic and small commercial fuel systems, the proper enforcement of the regulations, a public education campaign, an emergency response capability, tax incentives to consumers for expenditures associated with the upgrading or replacement of inadequate systems, support funding, the establishment of an emergency respond fund, and environmental cleanup requirements. figs

  20. Response to a fuel oil spill in the Albufera de Alcudia natural park on Mallorca Island

    Energy Technology Data Exchange (ETDEWEB)

    Bergueiro, J.R.; Moreno, S.; Guijarro, S.; Serra, F. [Universitat de les Illes Balears, Palma de Mallorca, Baleares (Spain); Perez-Navarro, A. [Universidad Politecnica de Valencia, Valencia (Spain); Kantin, R. [IFREMER, la Tremblade (France); Diez, E. [Transportes Salas Simo S.L., Palma de Mallorca (Spain)

    2002-07-01

    On June 12, 2001, a tanker spilled 14,500 liters of low sulphur fuel in a canal within an environmentally sensitive and ecologically rich, salt water lagoon of the Albufera de Alcudia Natural park on the island of Mallorca. Part of the contingency plan to minimize the impact of the spill included the use of a sorbent material on each side of the spill, followed by pumping the hydrocarbons out of the reed grasses, water, soil and sediments. The tanker was removed to avoid any further spill. The 428 tons of affected oil was moved by trucks to a temporary storage area in an adjacent lot where it was separated into 3 groups according to the treatment required. Polluted reed grass from a sugar cane plantation was mildly polluted. Another group was highly polluted, and the final group showed low level pollution. The fuel oil containing water, soil and sugar cane plantation material was analyzed to obtain average values of fuel oil per liter of water and fuel oil per kilogram of dry material. Material from the sugar cane plantation was burnt in an incinerator while the rest of the material was left to dry for 3 months before it was moved to an area for treatment in fenced containers designed with a slope for collecting leachates. Two 1.5 m deep wells were dug to accumulate the leachate. Analysis of the polluted reed grass samples one month after the spill indicated a concentration of 0.26 g of fuel oil per gram of dried reed grass which does not present any danger to flora and fauna. Observations made in September following the spill indicated a significant improvement in the state of reed grass and water within the affected area. The remediation effort was considered to be very efficient and total recovery of the affected area has been verified. 4 refs., 1 tab., 17 figs.

  1. Response to a fuel oil spill in the Albufera de Alcudia natural park on Mallorca Island

    International Nuclear Information System (INIS)

    Bergueiro, J.R.; Moreno, S.; Guijarro, S.; Serra, F.; Perez-Navarro, A.; Kantin, R.; Diez, E.

    2002-01-01

    On June 12, 2001, a tanker spilled 14,500 liters of low sulphur fuel in a canal within an environmentally sensitive and ecologically rich, salt water lagoon of the Albufera de Alcudia Natural park on the island of Mallorca. Part of the contingency plan to minimize the impact of the spill included the use of a sorbent material on each side of the spill, followed by pumping the hydrocarbons out of the reed grasses, water, soil and sediments. The tanker was removed to avoid any further spill. The 428 tons of affected oil was moved by trucks to a temporary storage area in an adjacent lot where it was separated into 3 groups according to the treatment required. Polluted reed grass from a sugar cane plantation was mildly polluted. Another group was highly polluted, and the final group showed low level pollution. The fuel oil containing water, soil and sugar cane plantation material was analyzed to obtain average values of fuel oil per liter of water and fuel oil per kilogram of dry material. Material from the sugar cane plantation was burnt in an incinerator while the rest of the material was left to dry for 3 months before it was moved to an area for treatment in fenced containers designed with a slope for collecting leachates. Two 1.5 m deep wells were dug to accumulate the leachate. Analysis of the polluted reed grass samples one month after the spill indicated a concentration of 0.26 g of fuel oil per gram of dried reed grass which does not present any danger to flora and fauna. Observations made in September following the spill indicated a significant improvement in the state of reed grass and water within the affected area. The remediation effort was considered to be very efficient and total recovery of the affected area has been verified. 4 refs., 1 tab., 17 figs

  2. Biodegradation of spilled diesel fuel in agricultural soil: Effect of humates, zeolite, and bioaugmentation

    Czech Academy of Sciences Publication Activity Database

    Kuráň, P.; Trögl, J.; Nováková, J.; Pilařová, V.; Dáňová, P.; Pavlorková, J.; Kozler, J.; Novák, František; Popelka, J.

    -, č. 642427 (2014) ISSN 1537-744X Grant - others:GA MPO(CZ) FR-TI1/456 Institutional support: RVO:60077344 Keywords : biodegradation * spilled diesel fuel * agricultural soil Subject RIV: DK - Soil Contamination ; De-contamination incl. Pesticides Impact factor: 1.219, year: 2013 http://dx.doi.org/10.1155/2014/642427

  3. Investigation of evaporation and biodegradation of fuel spills in Antarctica: II-extent of natural attenuation at Casey Station.

    Science.gov (United States)

    Snape, Ian; Ferguson, Susan H; Harvey, Paul McA; Riddle, Martin J

    2006-03-01

    In many temperate regions, fuel and oil spills are sometimes managed simply by allowing natural degradation to occur, while monitoring soils and groundwater to ensure that there is no off-site migration or on-site impact. To critically assess whether this approach is suitable for coastal Antarctic sites, we investigated the extent of evaporation and biodegradation at three old fuel spills at Casey Station. Where the contaminants migrated across frozen ground, probably beneath snow, approximately half the fuel evaporated in the first few months prior to infiltration at the beginning of summer. Once in the ground, however, evaporation rates were negligible. In contrast, minor spills from fuel drums buried in an abandoned waste disposal site did not evaporate to the same extent. Biodegradation within all three spill sites is generally very minor. We conclude that natural attenuation is not a suitable management strategy for fuel-contaminated soils in Antarctic coastal regions.

  4. The effectiveness testing of oil spill-treating agents

    International Nuclear Information System (INIS)

    Fingas, M.F.; Kyle, D.A.; Laroche, N.; Fieldhouse, B.; Sergy, G.; Stoodley, G.

    1995-01-01

    Laboratory effectiveness tests have been developed for four classes of oil spill treating agents: solidifiers, demulsifying agents, surface-washing agents and dispersants. Several treating agent products in these four categories have been tested for effectiveness. The aquatic toxicity of these agents is an important factor and has been measured for many products. These results are presented. Solidifiers or gelling agents solidify oil. Test results show that solidifiers require between 16% and 200% of agent by weight compared to the oil. De-emulsifying agents or emulsion breakers prevent the formation of or break water-in-oil emulsions. Surfactant-containing materials are of two types, surface-washing agents and dispersants. Testing has shown that effectiveness is orthogonal for these two types of treating agents. Tests of surface washing agents show that only a few agents have effectiveness of 25 to 55%, where this is defined as the percentage of oil removed from a test surface. Dispersant effectiveness results using the swirling flask test are reported. Heavy oils show effectiveness values of about 1%, medium crudes of about 10%, light crude oils of about 30% and very light oils of about 90%

  5. Spill-of-opportunity testing of dispersant effectiveness at the Mega Borg oil spill

    International Nuclear Information System (INIS)

    Payne, J.R.; Martrano, R.J.; Reilly, T.J.; Lindblom, G.P.; Kennicutt, M.C. II; Brooks, J.M.

    1993-01-01

    The release of 3.9 million gallons of Angola Planca crude oil from the stricken tanker Mega Borg 57 miles offshore of Galveston, Texas in June 1990 provided a valuable opportunity to document dispersant effectiveness under field conditions. Aerial application of Corexit 9527 (968 gallons total in four adjacent passes) onto an identified test portion of the slick was evaluated by concurrent observations from a command-and-control aircraft and surface vessels (with videotape and 35-mm photographic documentation) and ground truth measurements, including continuous 4-meter-depth ultraviolet/fluorescence and a discrete water sampling program. Using the study plan outlined by Payne and colleagues, target and control areas were designated before dispersant application by deployment of smoke bombs and coded three-meter drogues. Postdispersant surface vessel placement and 30 liter water sampling activities from the Texas A ampersand M research vessel HOS Citation were aided by the smoke bombs, the free-drifting drogues, and directions from the command-and-control aircraft. Subsequent FID GC and GC/MS analyses of water sample extracts allowed quantitation of the dispersed oil concentrations under both treated and control areas. Although the spilled oil was extremely light (API gravity 39.0) and subject to significant natural dispersion, the field observations, filmed documentation, and water column data clearly demonstrated an increase in dispersed oil concentrations beneath the treated slick. The distribution of dispersed oil droplets was very heterogeneous and reflected the patchy distribution of oil on the water surface before dispersant application. Maximum concentrations of dispersed hydrocarbons in the center of the treated zone were 22,000 μg/L (22 ppm) for total aliphatic and 5.6 μg/L (5.6 ppb) for total aromatics 60 to 90 minutes after dispersant application. Elevated levels were generally limited to the upper 1 to 3 meters of the water column

  6. Some recommendations for testing oil spill computer models

    International Nuclear Information System (INIS)

    Garcia-Martinez, R.

    1998-01-01

    According to a recent state-of-the-art review of modelling transport and fate of spills, more than 50 oil spill models have been developed in the last 30 years. Even though some of these models are used for spill response actions during accidents, environmental impact assessment, contingency planning and response training, there are no standard methodologies to evaluate their quality. This article presents some ideas that may contribute to design a set of standard benchmarks that would allow users and developers to assess models on a rational basis. (author)

  7. 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

  8. Experimental plans for LMFBR cavity liner sodium spill test LT-1

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Newell, G.A.

    1976-01-01

    Reinforced concrete is an important material of construction in LMFBR cavities and cells. Steel liners are often installed on the concrete surfaces to provide a gastight seal for minimizing air inleakage to inerted cell atmospheres and to protect the concrete from direct contact with sodium in the event of a sodium spill. In making safety assessment analyses, it is of interest to determine the adequacy of the liners to maintain their leaktightness during postulated accidents involving large sodium spills. However, data for basing analytical assessments of cell liners are very meager and an experimental program is underway at HEDL to provide some of the needed information. The HEDL cell liner evaluation program consists of both bench-scale feature tests and large-scale sodium spill demonstration tests. The plans for the first large-scale sodium spill test (LT-1) are the subject of this paper

  9. Effectiveness of bioremediation for the Prestige fuel spill : a summary of case studies

    International Nuclear Information System (INIS)

    Gallego, J.R.; Gonzalez-Rojas, E.; Pelaez, A.I.; Sanchez, J; Garcia-Martinez, M.J.; Llamas, J.F.

    2006-01-01

    This paper described novel bioremediation strategies used to remediate coastal areas in Spain impacted by the Prestige fuel oil spill in 2002. The bioremediation techniques were applied after hot pressurized water washing was used to remove hydrocarbons adhering to shorelines and rocks. Bioremediation strategies included monitored natural attenuation as well as accelerating biodegradation by stimulating indigenous populations through the addition of exogenous microbial populations. The sites selected for bioremediation were rocky shorelines of heterogenous granitic sediments with grain sizes ranging from sands to huge boulders; limestone-sandstone pebbles and cobbles; and fuel-coated limestone cliffs. Total surface area covered by the fuel was determined through the use of image analysis calculations. A statistical measurement of the fuel layer thickness was calculated by averaging the weights of multiple-fuel sampling increments. Bioremediation products included the use of oleophilic fertilizers; a biodegradable surfactant; and a microbial seeding agent. Determinations of saturate, aromatic, resins, and asphaltene (SARA) were performed using maltenes extraction and liquid chromatography. Microbial plating and selective enrichment with fuel as the sole carbon source were used to monitor the evolution of microbial populations in a variety of experiments. It was concluded that the biostimulation technique enhanced the efficiency of the in situ oleophilic fertilizers. 17 refs., 2 tabs., 6 figs

  10. Effectiveness of bioremediation for the Prestige fuel spill : a summary of case studies

    Energy Technology Data Exchange (ETDEWEB)

    Gallego, J.R. [Oviedo Univ., Asturias (Spain); Gonzalez-Rojas, E.; Pelaez, A.I.; Sanchez, J [Oviedo Univ., Asturias (Spain). Inst. de Biotecnologia de Asturias; Garcia-Martinez, M.J.; Llamas, J.F. [Univ. Polictenica de Madrid, Madrid (Spain). Laboratorio de Estratigrafia Biomolecular

    2006-07-01

    This paper described novel bioremediation strategies used to remediate coastal areas in Spain impacted by the Prestige fuel oil spill in 2002. The bioremediation techniques were applied after hot pressurized water washing was used to remove hydrocarbons adhering to shorelines and rocks. Bioremediation strategies included monitored natural attenuation as well as accelerating biodegradation by stimulating indigenous populations through the addition of exogenous microbial populations. The sites selected for bioremediation were rocky shorelines of heterogenous granitic sediments with grain sizes ranging from sands to huge boulders; limestone-sandstone pebbles and cobbles; and fuel-coated limestone cliffs. Total surface area covered by the fuel was determined through the use of image analysis calculations. A statistical measurement of the fuel layer thickness was calculated by averaging the weights of multiple-fuel sampling increments. Bioremediation products included the use of oleophilic fertilizers; a biodegradable surfactant; and a microbial seeding agent. Determinations of saturate, aromatic, resins, and asphaltene (SARA) were performed using maltenes extraction and liquid chromatography. Microbial plating and selective enrichment with fuel as the sole carbon source were used to monitor the evolution of microbial populations in a variety of experiments. It was concluded that the biostimulation technique enhanced the efficiency of the in situ oleophilic fertilizers. 17 refs., 2 tabs., 6 figs.

  11. Source identification of underground fuel spills by solid-phase microextraction/high-resolution gas chromatography/genetic algorithms.

    Science.gov (United States)

    Lavine, B K; Ritter, J; Moores, A J; Wilson, M; Faruque, A; Mayfield, H T

    2000-01-15

    Solid-phase microextraction (SPME), capillary column gas chromatography, and pattern recognition methods were used to develop a potential method for typing jet fuels so a spill sample in the environment can be traced to its source. The test data consisted of gas chromatograms from 180 neat jet fuel samples representing common aviation turbine fuels found in the United States (JP-4, Jet-A, JP-7, JPTS, JP-5, JP-8). SPME sampling of the fuel's headspace afforded well-resolved reproducible profiles, which were standardized using special peak-matching software. The peak-matching procedure yielded 84 standardized retention time windows, though not all peaks were present in all gas chromatograms. A genetic algorithm (GA) was employed to identify features (in the standardized chromatograms of the neat jet fuels) suitable for pattern recognition analysis. The GA selected peaks, whose two largest principal components showed clustering of the chromatograms on the basis of fuel type. The principal component analysis routine in the fitness function of the GA acted as an information filter, significantly reducing the size of the search space, since it restricted the search to feature subsets whose variance is primarily about differences between the various fuel types in the training set. In addition, the GA focused on those classes and/or samples that were difficult to classify as it trained using a form of boosting. Samples that consistently classify correctly were not as heavily weighted as samples that were difficult to classify. Over time, the GA learned its optimal parameters in a manner similar to a perceptron. The pattern recognition GA integrated aspects of strong and weak learning to yield a "smart" one-pass procedure for feature selection.

  12. Fuel Conservation by the Application of Spill Prevention and Failsafe Engineering (A Guideline Manual)

    Energy Technology Data Exchange (ETDEWEB)

    Goodier, J. Leslie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Water and Land Resources Department, Office of Marine and Environmental Engineering; Siclari, Robert J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Water and Land Resources Department, Office of Marine and Environmental Engineering; Garrity, Phyllis A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Water and Land Resources Department, Office of Marine and Environmental Engineering

    1980-10-30

    From a series of nationwide plant surveys dedicated to spill prevention, containment and countermeasure evaluation, coupled with spill response action activities, a need was determined for a spill prevention guideline manual. From Federally accumulated statistics for oil and hazardous substance spills, the authors culled information on spills of hydrocarbon products. In 1978, a total of 1456 oil spills were reported compared to 1451 in 1979. The 1978 spills were more severe, however, since 7,289,163 gallons of oil were accidentally discharged. In 1979, the gallons spilled was reduced to 3,663,473. These figures are derived from reported spills; it is highly possible that an equal amount was spilled and not reported. Spills effectively contained within a plant property that do not enter a navigational waterway need not be reported. Needless to say, there is a tremendous annual loss of oil products due to accidental spillage during transportation, cargo transfer, bulk storage and processing. As an aid to plant engineers and managers, Federal workers, fire marshalls and fire and casualty insurance inspectors, the document is offered as a spill prevention guide. The manual defines state-of-the-art spill prevention practices and automation techniques that can reduce spills caused by human error. Whenever practical, the cost of implementation is provided to aid equipment acquisition and installation budgeting. To emphasize the need for spill prevention activities, historic spills are briefly described after which remedial action is defined in an appropriate section of the manual. The section on plant security goes into considerable depth since to date no Federal agency or trade association has provided industry with guidelines on this important phase of plant operation. The intent of the document is to provide finger-tip reference material that can be used by interested parties in a nationwide effort to reduce loss of oil from preventable spills.

  13. Calibration and testing of IKU's oil spill contingency and response (OSCAR) model system

    International Nuclear Information System (INIS)

    Reed, M.; Aamo, O.M.; Downing, K.

    1996-01-01

    A computer modeling system entitled Oil Spill Contingency and Response (OSCAR), was calibrated and tested using a variety of field observations. The objective of the exercise was to establish model credibility and increase confidence in efforts to compare alternate oil spill response strategies, while maintaining a balance between response costs and environmental protection. The key components of the system are IKU's data-based oil weathering model, a three dimensional oil trajectory and chemical fates model, an oil spill combat model, and exposure models for fish, ichthyoplankton, birds, and marine mammals. Most modelled calculations were in good agreement with field observations. One discrepancy was found which could be attributed to an underestimation of wind drift in the current model. 21 refs., 4 tabs., 32 figs

  14. Spills Action Centre summary report of 1992 spills

    International Nuclear Information System (INIS)

    1993-11-01

    Environment Ontario's Spills Action Center (SAC) receives and initiates response to spills and other urgent environmental incidents on a 24 h per day basis. The center documented 14,588 occurrence reports in 1992. Two thirds of these involved a range of ministry notification requirements and environmental complaints, while one third involved spills. Information on spills reported in 1992 are summarized. The 5,014 spills reported to SAC in 1992 represent a 5% decrease from 1991. Fewer spills to air accounted for this decrease, while the number of spills to land and water remained unchanged. Oil and fuel spills accounted for 59% of spilled material. Chemical or chemical solutions accounted for ca 15%, wastes or wastewaters 18%, gaseous emissions 6% and unknown for the remainder. Around 20% of spills were less than 10 liters, 57% were less than 100 liters, and 86% were less than 1000 liters. About 28% of the spills had a confirmed environmental impact or adverse affect, two thirds involving soil contamination and around one fifth involving surface water contamination. Twenty-three spills resulted in human health and safety concerns. Around 45% of all spills were completely cleaned up, and an additional 22% were partially cleaned up. Industrial sectors with the largest proportion of reported spills were: transportation, 16%; petroleum, 13%; metallurgical, 6%; general manufacturing, 5%; and chemical, 5%. Public sector spills accounted for 18% of reported spills. Motor vehicles were the largest sources of spills accounting for over 28% of reported spills. 14 figs., 14 tabs

  15. 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

  16. 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.

  17. A Study of Spill Control Characteristics of JP-8 and Conventional Diesel Fuel with a Common Rail Direct Injection System

    Directory of Open Access Journals (Sweden)

    Seomoon Yang

    2017-12-01

    Full Text Available Diversification of energy sources is a key task for decreasing environmental impacts and global emission of gases. JP-8, a fuel derived from natural gas, coal, biomass, and waste plastics, is a bright prospect. JP-8 is considered a multi-source multi-purpose fuel, with several applications. A preliminary characterization of the JP-8 injection rate and injection quantity behavior was investigated based on the high-pressure common rail injection system used in a heavy-duty engine. According to the spill injection and injection pressure, a trade-off trend between injection rate and injection quantity was observed. As expected, pilot injection of JP-8 aviation fuel and diesel fuel affects the spray quantity and injection evolution of the subsequent operation without pilot injection. The difference in spilling between diesel and JP-8 aviation fuel is greater than the difference in injection amount per time; in the process of controlling the injector solenoid through ECU (Electric Control Units, the oil pressure valve and the needle valve operate to a higher extent in order to maintain the diesel fuel’s injection quantity volume. It was found that the total injection quantity was decreased by adding 20% pilot injection duration. Because the pilot injection quantity causes solenoid response, loss and needle lift stroke friction loss.

  18. Assessing fuel spill risks in polar waters: Temporal dynamics and behaviour of hydrocarbons from Antarctic diesel, marine gas oil and residual fuel oil.

    Science.gov (United States)

    Brown, Kathryn E; King, Catherine K; Kotzakoulakis, Konstantinos; George, Simon C; Harrison, Peter L

    2016-09-15

    As part of risk assessment of fuel oil spills in Antarctic and subantarctic waters, this study describes partitioning of hydrocarbons from three fuels (Special Antarctic Blend diesel, SAB; marine gas oil, MGO; and intermediate grade fuel oil, IFO 180) into seawater at 0 and 5°C and subsequent depletion over 7days. Initial total hydrocarbon content (THC) of water accommodated fraction (WAF) in seawater was highest for SAB. Rates of THC loss and proportions in equivalent carbon number fractions differed between fuels and over time. THC was most persistent in IFO 180 WAFs and most rapidly depleted in MGO WAF, with depletion for SAB WAF strongly affected by temperature. Concentration and composition remained proportionate in dilution series over time. This study significantly enhances our understanding of fuel behaviour in Antarctic and subantarctic waters, enabling improved predictions for estimates of sensitivities of marine organisms to toxic contaminants from fuels in the region. Copyright © 2016 Elsevier Ltd. All rights reserved.

  19. 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.

  20. A comparative assessment of the fate and effects of similar Orimulsion and heavy fuel oil spills in the Milford Haven Estuary, UK

    International Nuclear Information System (INIS)

    Mendelsohn, D.L.; Rines, H.M.; Christensen, F.T.; Isajil, T.; French, D.; Edwards, N.

    1996-01-01

    A Spill Impact Assessment Program (SIMAP) was used to simulate spills of 1,000 metric tons each of No. 6 fuel oil and Orimulsion into the Milford Haven Estuary. The experiments were conducted because many oil burning power generating stations worldwide have considered the use of Orimulsion (a mixture of 70% bitumen, 30% water and a surfactant) in place of the conventional heavy fuel oil No. 6. An assessment of the potential effects of an accidental spill of Orimulsion was therefore conducted. A comparison of that spill to one of conventional fuel oil was part of the evaluation. The experiments were run for three weeks, each under identical environmental conditions and impacts on the local habitats and biota were calculated. The model predicted that a spill of No. 6 fuel (which floats because of its low density relative to sea water) would result in the loss of about 14% of the total bird population. Fish kills would be very small. A spill of Orimulsion (which readily disperses into the water column because of the presence of a surfactant) would result in the loss of about 0.4% of the adult fish and 3.7% of young fish, while birds losses would be negligible. 22 refs., 3 tabs., 12 figs

  1. Evaluation of three oil spill laboratory dispersant effectiveness tests

    International Nuclear Information System (INIS)

    Sullivan, D.; Farlow, J.; Sahatjian, K.A.

    1993-01-01

    Chemical dispersants can be used to reduce the interfacial tension of floating oil slicks so that the oils disperse more rapidly into the water column and thus pose less of a threat to shorelines, birds, and marine mammals. The laboratory test currently specified in federal regulations to measure dispersant effectiveness is not especially easy or inexpensive, and generates a rather large quantity of oily waste water. This paper describes the results of an effort by the EPA to identify a more suitable laboratory dispersant effectiveness test. EPA evaluated three laboratory methods: the Revised Standard Dispersant Effectiveness Test currently used (and required by regulation) in the United States, the swirling flask test (developed by Environment Canada), and the IFP-dilution test (used in france and other European countries). Six test oils and three dispersants were evaluated; dispersants were applied to the oil at an average 1:10 ratio (dispersant to oil) for each of the three laboratory methods. Screening efforts were used to focus on the most appropriate oil/dispersant combination for detailed study. A screening criterion was established that required a combination that gave at least 20% effectiveness results. The selected combination turned out to be Prudhoe Bay crude oil and the dispersant Corexit 9527. This combination was also most likely to be encountered in US coastal waters. The EPA evaluation concluded that the three tests gave similar precision results, but that the swirling flask test was fastest, cheapest, simplest, and required least operator skill. Further, EPA is considering conducting the dispersant effectiveness test itself, rather than having data submitted by a dispersant manufacturer, and establishing an acceptability criterion (45% efficiency) which would have to be met before a dispersant could be placed on the Product Schedule of the National Contingency Plan (NCP)

  2. Migracija dizel goriva izlivenog u slojeve zemljišta / Migration of diesel fuel spilled in subsurface layers of soil

    Directory of Open Access Journals (Sweden)

    Mladen Vuruna

    2005-09-01

    Full Text Available U radu su prikazane osnovne fizičko-hemijske karakteristike dizel goriva i zemljišta. Objašnjena je migracija izlivenog naftnog zagađivača kroz vertikalni profil zemljišta. U eksperimentalnom delu ispitivane su koncentracije dizel goriva i relativne koncentracije n-alkana u površinskim slojevima peska, u koje gorivo dospeva kao posledica akcidentnog izlivanja. Utvrđeno je da se koncentracije dizel goriva menjaju sa vremenom nakon izlivanja u svim ispitivanim slojevima. Takođe, utvrđeno je da se dizel gorivo, kao potencijalni zagađivač, u prvih šest nedelja, uglavnom, zadržava u površinskom sloju dubine 30 cm, a objašnjene su i mogućnosti sanacije zagađenog zemljišta. / The basic physical and chemical properties of both diesel fuel and soil have been given in this article and oil pollutants migration through vertical soil profile have been explained as well. In the experimental part of the paper both the concentrations of diesel fuel and relative concentrations of n-alkynes spilled in sandy soil by accident have been investigated. It has been proven that the concentrations of diesel fuel have changed in all layers of soil depending on the time after spill. Diesel fuel as possible pollutant has been retained 30 cm deep in sandy soil during six weeks after spill. Finally, cleanup techniques of polluted soil have been explained.

  3. 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...

  4. Coconut endocarp and mesocarp as both biosorbents of dissolved hydrocarbons in fuel spills and as a power source when exhausted.

    Science.gov (United States)

    Luis-Zarate, Victor Hugo; Rodriguez-Hernandez, Mayra Cecilia; Alatriste-Mondragon, Felipe; Chazaro-Ruiz, Luis Felipe; Rangel-Mendez, Jose Rene

    2018-04-01

    Health and environmental problems associated with the presence of toxic aromatic compounds in water from oil spills have motivated research to develop effective and economically viable strategies to remove these pollutants. In this work, coconut shell (endocarp), coconut fiber (mesocarp) and coconut shell with fiber (endocarp and mesocarp) obtained from coconut (Cocos nucifera) waste were evaluated as biosorbents of benzene, toluene and naphthalene from water, considering the effect of the solution pH (6-9) and the presence of dissolved organic matter (DOM) in natural water (14 mg/L). In addition, the heat capacity of saturated biosorbents was determined to evaluate their potential as an alternative power source to conventional fossil fuels. Tests of N 2 physisorption, SEM, elemental and fiber analysis, ATR-FTIR and acid-based titrations were performed in order to understand the materials' characteristics, and to elucidate the biosorbents' hydrocarbon adsorption mechanism. Coconut fiber showed the highest adsorption capacities (222, 96 and 5.85 mg/g for benzene, toluene and naphthalene, respectively), which was attributed to its morphologic characteristics and to its high concentration of phenolic groups, associated with the lignin structure. The pH of the solution did not have a significant influence on the removal of the contaminants, and the presence of DOM improved the adsorption capacities of aromatic hydrocarbons. The adsorption studies showed biphasic isotherms, which highlighted the strong affinity between the molecules adsorbed on the biosorbents and the aromatic compounds remaining in the solution. Finally, combustion heat analysis of coconut waste saturated with soluble hydrocarbons showed that the heat capacity increased from 4407.79 cal/g to 5064.43 ± 11.6 cal/g, which is comparable with that of woody biomass (3400-4000 cal/g): this waste biomass with added value could be a promising biofuel. Copyright © 2018 Elsevier Ltd. All rights

  5. Development of a field testing protocol for identifying Deepwater Horizon oil spill residues trapped near Gulf of Mexico beaches

    Science.gov (United States)

    Han, Yuling

    2018-01-01

    The Deepwater Horizon (DWH) accident, one of the largest oil spills in U.S. history, contaminated several beaches located along the Gulf of Mexico (GOM) shoreline. The residues from the spill still continue to be deposited on some of these beaches. Methods to track and monitor the fate of these residues require approaches that can differentiate the DWH residues from other types of petroleum residues. This is because, historically, the crude oil released from sources such as natural seeps and anthropogenic discharges have also deposited other types of petroleum residues on GOM beaches. Therefore, identifying the origin of these residues is critical for developing effective management strategies for monitoring the long-term environmental impacts of the DWH oil spill. Advanced fingerprinting methods that are currently used for identifying the source of oil spill residues require detailed laboratory studies, which can be cost-prohibitive. Also, most agencies typically use untrained workers or volunteers to conduct shoreline monitoring surveys and these worker will not have access to advanced laboratory facilities. Furthermore, it is impractical to routinely fingerprint large volumes of samples that are collected after a major oil spill event, such as the DWH spill. In this study, we propose a simple field testing protocol that can identify DWH oil spill residues based on their unique physical characteristics. The robustness of the method is demonstrated by testing a variety of oil spill samples, and the results are verified by characterizing the samples using advanced chemical fingerprinting methods. The verification data show that the method yields results that are consistent with the results derived from advanced fingerprinting methods. The proposed protocol is a reliable, cost-effective, practical field approach for differentiating DWH residues from other types of petroleum residues. PMID:29329313

  6. Development of a field testing protocol for identifying Deepwater Horizon oil spill residues trapped near Gulf of Mexico beaches.

    Science.gov (United States)

    Han, Yuling; Clement, T Prabhakar

    2018-01-01

    The Deepwater Horizon (DWH) accident, one of the largest oil spills in U.S. history, contaminated several beaches located along the Gulf of Mexico (GOM) shoreline. The residues from the spill still continue to be deposited on some of these beaches. Methods to track and monitor the fate of these residues require approaches that can differentiate the DWH residues from other types of petroleum residues. This is because, historically, the crude oil released from sources such as natural seeps and anthropogenic discharges have also deposited other types of petroleum residues on GOM beaches. Therefore, identifying the origin of these residues is critical for developing effective management strategies for monitoring the long-term environmental impacts of the DWH oil spill. Advanced fingerprinting methods that are currently used for identifying the source of oil spill residues require detailed laboratory studies, which can be cost-prohibitive. Also, most agencies typically use untrained workers or volunteers to conduct shoreline monitoring surveys and these worker will not have access to advanced laboratory facilities. Furthermore, it is impractical to routinely fingerprint large volumes of samples that are collected after a major oil spill event, such as the DWH spill. In this study, we propose a simple field testing protocol that can identify DWH oil spill residues based on their unique physical characteristics. The robustness of the method is demonstrated by testing a variety of oil spill samples, and the results are verified by characterizing the samples using advanced chemical fingerprinting methods. The verification data show that the method yields results that are consistent with the results derived from advanced fingerprinting methods. The proposed protocol is a reliable, cost-effective, practical field approach for differentiating DWH residues from other types of petroleum residues.

  7. 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

  8. 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

  9. Monitoring of Olympic National Park Beaches to determine fate and effects of spilled bunker C fuel oil

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.A.; Cullinan, V.I.; Crecelius, E.A.; Fortman, T.J.; Citterman, R.J.; Fleischmann, M.L.

    1990-10-01

    On December 23, 1988, the barge Nestucca was accidentally struck by its tow, a Souse Brothers Towing Company tug, releasing approximately 230,000 gallons of Bunker C fuel oil and fouling beaches from Grays Harbor north to Vancouver Island. Affected beaches in Washington included a 40-mile-long strip that has been recently added to Olympic National Park. The purpose of the monitoring program documented in this report was to determine the fate of spilled Bunker C fuel oil on selected Washington coastal beaches. We sought to determine (1) how much oil remained in intertidal and shallow subtidal habitats following clean-up and weathering, (2) to what extent intertidal and/or shallow subtidal biotic assemblages have been contaminated, and (3) how rapidly the oil has left the ecosystem. 45 refs., 18 figs., 8 tabs.

  10. Spills Action Centre summary report of 1995 spills

    International Nuclear Information System (INIS)

    1996-12-01

    A summary of spills reported to the Ontario Ministry of Environment and Energy during 1995 was presented. The Ministry's Spill Action Centre is on call 24-hours per day to receive and respond to reports of spills and other urgent environmental incidents. Some 5,000 spills were documented in 1995. Oils and fuels accounted for 59 per cent of the spilled materials, chemicals and chemical solutions for 17 per cent, wastes and waste waters for 16 per cent, and gaseous materials for 5 per cent. Unknown materials accounted for 3 per cent. Most of the spills involved small volumes. Equipment failure and operator error were the major reasons for spills. All occurrences reported are stored on a computerized database. The information is used to develop new pollution abatement programs and spill prevention initiatives as trends are identified. 14 tabs., 14 figs

  11. TESTING THE GENERALIZATION EFFICIENCY OF OIL SLICK CLASSIFICATION ALGORITHM USING MULTIPLE SAR DATA FOR DEEPWATER HORIZON OIL SPILL

    Directory of Open Access Journals (Sweden)

    C. Ozkan

    2012-07-01

    Full Text Available Marine oil spills due to releases of crude oil from tankers, offshore platforms, drilling rigs and wells, etc. are seriously affecting the fragile marine and coastal ecosystem and cause political and environmental concern. A catastrophic explosion and subsequent fire in the Deepwater Horizon oil platform caused the platform to burn and sink, and oil leaked continuously between April 20th and July 15th of 2010, releasing about 780,000 m3 of crude oil into the Gulf of Mexico. Today, space-borne SAR sensors are extensively used for the detection of oil spills in the marine environment, as they are independent from sun light, not affected by cloudiness, and more cost-effective than air patrolling due to covering large areas. In this study, generalization extent of an object based classification algorithm was tested for oil spill detection using multiple SAR imagery data. Among many geometrical, physical and textural features, some more distinctive ones were selected to distinguish oil and look alike objects from each others. The tested classifier was constructed from a Multilayer Perception Artificial Neural Network trained by ABC, LM and BP optimization algorithms. The training data to train the classifier were constituted from SAR data consisting of oil spill originated from Lebanon in 2007. The classifier was then applied to the Deepwater Horizon oil spill data in the Gulf of Mexico on RADARSAT-2 and ALOS PALSAR images to demonstrate the generalization efficiency of oil slick classification algorithm.

  12. 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

  13. In situ bioremediation of a diesel fuel spill in northern Manitoba

    International Nuclear Information System (INIS)

    Hryhoruk, C.D.

    1994-01-01

    At a northern Manitoba airport, a site was contaminated with diesel fuel, which was confined within the unsaturated zone in silt and silty sand. A two-phase bioremediation process was designed and implemented in-situ in a pilot test. The first phase, ground surface spraying, involved mixing nutrients (ammonium-nitrogen and orthophosphate) with water in a tank and then spraying the mixture on the ground surface above the diesel plume. The second phase, a pump-cycle system, involved pumping groundwater from below the diesel plume into one of two tanks in series. The groundwater underwent both nutrient addition (weekly) and aeration in the tanks, then it was pumped into eight feeder wells which circumscribed an extraction well. Soil testing revealed that both remediation processes aided in increasing subsurface nutrient concentrations and the moisture content within the diesel plume. In addition, high total coliform counts were observed in both the silt and silty sand layers. This implied that conditions for suitable bioremediation can be developed in relatively fine-grained soil. Intermittent soil sampling at three locations over a 14-month period revealed that the diesel plume decreased in size by ca 30% and contaminant concentrations (diesel fuel) also decreased. Plume movement also occurred. The pump-cycle system remains operational. 67 refs., 77 figs., 9 tabs

  14. In situ bioremediation of a diesel fuel spill in northern Manitoba

    Energy Technology Data Exchange (ETDEWEB)

    Hryhoruk, C D

    1994-01-01

    At a northern Manitoba airport, a site was contaminated with diesel fuel, which was confined within the unsaturated zone in silt and silty sand. A two-phase bioremediation process was designed and implemented in-situ in a pilot test. The first phase, ground surface spraying, involved mixing nutrients (ammonium-nitrogen and orthophosphate) with water in a tank and then spraying the mixture on the ground surface above the diesel plume. The second phase, a pump-cycle system, involved pumping groundwater from below the diesel plume into one of two tanks in series. The groundwater underwent both nutrient addition (weekly) and aeration in the tanks, then it was pumped into eight feeder wells which circumscribed an extraction well. Soil testing revealed that both remediation processes aided in increasing subsurface nutrient concentrations and the moisture content within the diesel plume. In addition, high total coliform counts were observed in both the silt and silty sand layers. This implied that conditions for suitable bioremediation can be developed in relatively fine-grained soil. Intermittent soil sampling at three locations over a 14-month period revealed that the diesel plume decreased in size by ca 30% and contaminant concentrations (diesel fuel) also decreased. Plume movement also occurred. The pump-cycle system remains operational. 67 refs., 77 figs., 9 tabs.

  15. 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

  16. Steamlined Approach for Environmental Restoration (SAFER) Plan For Corrective Action Unit 394: Areas 12, 18, and 29, Spill/Release Sites, Nevada Test Site, Nevada (November 2001, Rev. 0)

    Energy Technology Data Exchange (ETDEWEB)

    U.S. Department of Energy, National Nuclear Security Administration Nevada Operations Office (NNSA/NV)

    2001-09-24

    This plan addresses the actions necessary for the characterization and closure of Corrective Action Unit (CAU) 394: Areas 12, 18, and 29, Spill/Release Sites, identified in the Federal Facility Agreement and Consent Order (FFACO). The CAU, located on the Nevada Test Site, consists of six Corrective Action Sites (CASs): CAS 12-25-04, UST 12-16-2 Waste Oil Release; CAS 18-25-02, Oil Spills; CAS 18-25-02, Oil Spills; CAS 18-25-03, Oil Spill; CAS 18-25-04, Spill (Diesel Fuel); CAS 29-44-01, Fuel Spill (a & b). Process knowledge is the basis for the development of the conceptual site models (CSMs). The CSMs describe the most probable scenario for current conditions at each site, and define the assumptions that are the basis for the SAFER plan. The assumptions are formulated from historical information and process knowledge. Vertical migration of contaminant(s) of potential concern (COPCs) is expected to be predominant over lateral migration in the absence of any barrier (with asphalt /concrete being the exception at least two of the CASs). Soil is the impacted or potentially impacted media at all the sites, with asphalt and/or concrete potentially impacted at two of the CASs. Radionuclides are not expected at any CAS; hydrocarbons are the primary COPC at each CAS, and can be used to guide the investigation; future land-use scenarios limit use to various nonresidential uses; and exposure scenarios are limited by future land-use scenarios to site workers. There is sufficient information and process knowledge from historical documentation regarding the expected nature and extent of potential contaminants to recommend closure of CAU 394 using the SAFER process. On completion of the field activities, a Closure Report will be prepared and submitted to the NDEP for review and approval.

  17. Steamlined Approach for Environmental Restoration (SAFER) Plan For Corrective Action Unit 394: Areas 12, 18, and 29, Spill/Release Sites, Nevada Test Site, Nevada (November 2001, Rev. 0); FINAL

    International Nuclear Information System (INIS)

    2001-01-01

    This plan addresses the actions necessary for the characterization and closure of Corrective Action Unit (CAU) 394: Areas 12, 18, and 29, Spill/Release Sites, identified in the Federal Facility Agreement and Consent Order (FFACO). The CAU, located on the Nevada Test Site, consists of six Corrective Action Sites (CASs): CAS 12-25-04, UST 12-16-2 Waste Oil Release; CAS 18-25-02, Oil Spills; CAS 18-25-02, Oil Spills; CAS 18-25-03, Oil Spill; CAS 18-25-04, Spill (Diesel Fuel); CAS 29-44-01, Fuel Spill (a and b). Process knowledge is the basis for the development of the conceptual site models (CSMs). The CSMs describe the most probable scenario for current conditions at each site, and define the assumptions that are the basis for the SAFER plan. The assumptions are formulated from historical information and process knowledge. Vertical migration of contaminant(s) of potential concern (COPCs) is expected to be predominant over lateral migration in the absence of any barrier (with asphalt /concrete being the exception at least two of the CASs). Soil is the impacted or potentially impacted media at all the sites, with asphalt and/or concrete potentially impacted at two of the CASs. Radionuclides are not expected at any CAS; hydrocarbons are the primary COPC at each CAS, and can be used to guide the investigation; future land-use scenarios limit use to various nonresidential uses; and exposure scenarios are limited by future land-use scenarios to site workers. There is sufficient information and process knowledge from historical documentation regarding the expected nature and extent of potential contaminants to recommend closure of CAU 394 using the SAFER process. On completion of the field activities, a Closure Report will be prepared and submitted to the NDEP for review and approval

  18. 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

  19. Actual directions in study of ecological consequences of a highly toxic 1,1-dimethylhydrazine-based rocket fuel spills

    Directory of Open Access Journals (Sweden)

    Bulat Kenessov

    2012-05-01

    Full Text Available The paper represents a review of the actual directions in study of ecological consequences of highly toxic 1,1-dimethylhydrazine-based rocket fuel spills. Recent results on study of processes of transformation of 1,1-dimethylhydrazine, identification of its main metabolites and development of analytical methods for their determination are generalized. Modern analytical methods of determination of 1,1-dimethylhydrazine and its transformation products in environmental samples are characterized. It is shown that in recent years, through the use of most modern methods of physical chemical analysis and sample preparation, works in this direction made significant progress and contributed to the development of studies in adjacent areas. A character of distribution of transformation products in soils of fall places of first stages of rocket-carriers is described and the available methods for their remediation are characterized.

  20. Oil spills

    International Nuclear Information System (INIS)

    Katsouros, M.H.

    1992-01-01

    The world annually transports 1.7 billion tons of oil by sea, and oil spills, often highly concentrated discharges, are increasing from a variety of sources. The author discusses sources of oils spills: natural; marine transportation; offshore oil production; atmospheric sources; municipal industrial wastes and runoff. Other topics include: the fate of the spilled oil; the effects of the oil; the response to oil spills; and prevention of oil spills. 30 refs., 1 fig., 4 tabs

  1. 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

  2. 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.

  3. Evaluating a protocol for testing fire-resistant oil-spill containment boom

    International Nuclear Information System (INIS)

    Walton, W.D.; Twilley, W.H.; Hiltabrand, R.R.; Mullin, J.V.

    1998-01-01

    A series of experiments were conducted to evaluate a protocol for testing the ability of fire-resistant booms to withstand both fire and waves. Most response plans for in situ burning of oil at sea require the use of a fire-resistant boom to contain the oil during a burn. For this study, a wave tank was designed and constructed to assess the capabilities of a 15 m section of a boom subjected to a 5 m diameter fire with 0.15 m high waves. Five typical fire-resistant oil-spill containment booms were tested. The purpose of the project was to evaluate the test procedure, therefore the overall performance of the boom was not evaluated on a pass-fail criterion. The two most important aspects of the test method were repeatability and reproducibility. Some of the parameters tested included the effect of wind, waves, fire size, and fire duration. Methods to constrain the booms were also tested. 7 refs., 6 tabs., 7 figs

  4. Real-time petroleum spill detection system

    International Nuclear Information System (INIS)

    Dakin, D.T.

    2001-01-01

    A real-time autonomous oil and fuel spill detection system has been developed to rapidly detect of a wide range of petroleum products floating on, or suspended in water. The system consists of an array of spill detection buoys distributed within the area to be monitored. The buoys are composed of a float and a multispectral fluorometer, which looks up through the top 5 cm of water to detect floating and suspended petroleum products. The buoys communicate to a base station computer that controls the sampling of the buoys and analyses the data from each buoy to determine if a spill has occurred. If statistically significant background petroleum levels are detected, the system raises an oil spill alarm. The system is useful because early detection of a marine oil spill allows for faster containment, thereby minimizing the contaminated area and reducing cleanup costs. This paper also provided test results for biofouling, various petroleum product detection, water turbidity and wave tolerance. The technology has been successfully demonstrated. The UV light source keeps the optic window free from biofouling, and the electronics are fully submerged so there is no risk that the unit could ignite the vapours of a potential oil spill. The system can also tolerate moderately turbid waters and can therefore be used in many rivers, harbours, water intakes and sumps. The system can detect petroleum products with an average thickness of less than 3 micrometers floating on the water surface. 3 refs., 15 figs

  5. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 330: Areas 6, 22, and 23 Tanks and Spill Sites, Nevada Test Site, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    T. M. Fitzmaurice

    2001-08-01

    This Streamlined Approach for Environmental restoration (SAFER) plan addresses the action necessary for the closure of Corrective Action Unit (CAU) 330, Areas 6,22, and 23 Tanks and Spill Sites. The CAUs are currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO). This CAU is located at the Nevada Test Site (NTS) (Figure 1). CAU 330 consists of the following Corrective Action Sites (CASs): (1) CAS 06-02-04 - Consists of an underground tank and piping. This CAS is close to an area that was part of the Animal Investigation Program (AIP), conducted under the U.S. Public Health Service. Its purpose was to study and perform tests on the cattle and wild animals in and around the NTS that were exposed to radionuclides. It is unknown if this tank was part of these operations. (2) CAS 22-99-06 - Is a fuel spill that is believed to be a waste oil release which occurred when Camp Desert Rock was an active facility. This CAS was originally identified as being a small depression where liquids were poured onto the ground, located on the west side of Building T-1001. This building has been identified as housing a fire station, radio station, and radio net remote and telephone switchboard. (3) CAS 23-01-02 - Is a large aboveground storage tank (AST) farm that was constructed to provide gasoline and diesel storage in Area 23. The site consists of two ASTs, a concrete foundation, a surrounding earthen berm, associated piping, and unloading stations. (4) CAS 23-25-05 - Consists of an asphalt oil spill/tar release that contains a wash covered with asphalt oil/tar material, a half buried 208-liter (L) (55-gallon [gal]) drum, rebar, and concrete located in the vicinity.

  6. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 330: Areas 6, 22, and 23 Tanks and Spill Sites, Nevada Test Site, Nevada; TOPICAL

    International Nuclear Information System (INIS)

    T. M. Fitzmaurice

    2001-01-01

    This Streamlined Approach for Environmental restoration (SAFER) plan addresses the action necessary for the closure of Corrective Action Unit (CAU) 330, Areas 6,22, and 23 Tanks and Spill Sites. The CAUs are currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO). This CAU is located at the Nevada Test Site (NTS) (Figure 1). CAU 330 consists of the following Corrective Action Sites (CASs): (1) CAS 06-02-04 - Consists of an underground tank and piping. This CAS is close to an area that was part of the Animal Investigation Program (AIP), conducted under the U.S. Public Health Service. Its purpose was to study and perform tests on the cattle and wild animals in and around the NTS that were exposed to radionuclides. It is unknown if this tank was part of these operations. (2) CAS 22-99-06 - Is a fuel spill that is believed to be a waste oil release which occurred when Camp Desert Rock was an active facility. This CAS was originally identified as being a small depression where liquids were poured onto the ground, located on the west side of Building T-1001. This building has been identified as housing a fire station, radio station, and radio net remote and telephone switchboard. (3) CAS 23-01-02 - Is a large aboveground storage tank (AST) farm that was constructed to provide gasoline and diesel storage in Area 23. The site consists of two ASTs, a concrete foundation, a surrounding earthen berm, associated piping, and unloading stations. (4) CAS 23-25-05 - Consists of an asphalt oil spill/tar release that contains a wash covered with asphalt oil/tar material, a half buried 208-liter (L) (55-gallon[gal]) drum, rebar, and concrete located in the vicinity

  7. Air-deployable oil spill sampling devices review phase 2 testing. Volume 1

    International Nuclear Information System (INIS)

    Hawke, L.; Dumouchel, A.; Fingas, M.; Brown, C.E.

    2007-01-01

    SAIC Canada tested air deployable oil sampling devices for the Emergencies Science and Technology Division of Environment Canada in order to determine the applicability and status of these devices. The 3 devices tested were: Canada's SABER (sampling autonomous buoy for evidence recovery), the United States' POPEIE (probe for oil pollution evidence in the environment); and, Sweden's SAR Floatation 2000. They were tested for buoyancy properties, drift behaviour and sampler sorbent pickup ratios. The SAR and SABER both had lesser draft and greater freeboard, while the POPEIE had much greater draft than freeboard. All 3 devices could be used for oil sample collection in that their drift characteristics would allow for the SABER and SAR devices to be placed upwind of the slick while the POPEIE device could be placed downwind of an oil spill. The sorbent testing revealed that Sefar sorbent and Spectra sorbent used in the 3 devices had negative pickup ratios for diesel but performance improved as oil viscosity increased. Both sorbents are inert and capable of collecting oil in sufficient volumes for consistent fingerprinting analysis. 10 refs., 8 tabs., 8 figs

  8. 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.

  9. Oil Spill Cleanup

    Science.gov (United States)

    Kauble, Christena Ann

    2011-01-01

    Several classroom activities using a model of a seashore and an oil spill demonstrate the basic properties of oil spills in oceans. Students brainstorm about how to best clean up the mess. They work in teams, and after agreeing on how they will proceed, their method is tested by measuring the amount of oil removed and by rating the cleanliness of…

  10. Controlled Cold Helium Spill Test in the LHC Tunnel at CERN

    Science.gov (United States)

    Koettig, T.; Casas-Cubillos, J.; Chorowski, M.; Dufay-Chanat, L.; Grabowski, M.; Jedrusyna, A.; Lindell, G.; Nonis, M.; Vauthier, N.; van Weelderen, R.; Winkler, T.; Bremer, J.

    The helium cooled magnets of the LHC particle accelerator are installed in a confined space, formed by a 27 km circumference 3.8 m diameter underground tunnel. The vacuum enclosures of the superconducting LHC magnets are protected by a lift plate against excessive overpressure created by eventual leaks from the magnet helium bath, or from the helium supply headers. A three-meter long no stay zone has been defined centered to these plates, based on earlier scale model studies, to protect the personnel against the consequences of an eventual opening of such a lift plate. More recently several simulation studies have been carried out modelling the propagation of the resulting helium/air mixture along the tunnel in case of such a cold helium release at a rate in the range of 1 kg/s. To validate the different scale models and simulation studies, real life mock-up tests have been performed in the LHC, releasing about 1000 liter of liquid helium under standard operational tunnel conditions. Data recorded during these tests include oxygen level, temperature and flow speed as well as video recordings, taken up- and downstream of the spill point (-100 m to +200 m) with respect to the ventilation direction in the LHC tunnel. The experimental set-up and measurement results are presented. Generic effects found during the tests will be discussed to allow the transposal to possible cold helium release cases in similar facilities.

  11. Oil Spills

    Science.gov (United States)

    Oil spills often happen because of accidents, when people make mistakes or equipment breaks down. Other causes include natural disasters or deliberate acts. Oil spills have major environmental and economic effects. Oil ...

  12. 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.)

  13. 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.

  14. 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...

  15. 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...

  16. Bioremediation of oil spills

    International Nuclear Information System (INIS)

    Lynn, J.

    2001-01-01

    The conversion of oil to environmentally benign chemicals such as water and carbon dioxide by 'hydrocarbon-eating' bacteria is described. The emphasis is on a new process to selectively increase the population of 'oil eating' bacteria, a development that became the foundation for the second-generation bioremediation accelerator, Inipol EAP-22. Second-generation bioremediation products focus on providing nitrogen and phosphorus, chemicals that are not present in crude oil in readily available form, but are essential for the synthesis of proteins, nucleic acids, phospholipids and the energy metabolism of the bacteria. Providing these chemicals in the proper amounts encourages the preferential growth of oil-degrading microbes already present in the local biomass, thus overcoming the major limiting factor for biodegradation. These second-generation bioremediation products also have strong oleophilic properties engineered into them, to assure that the nutrients essential for the bacteria are in contact with the oil. The first major test for second-generation bioremediation accelerators came with the clean-up of the oil spill from the Exxon Valdez, a disaster that contaminated more than 120 kilometres of Alaskan beaches along the shores of Prince William Sound. The Inipol EAP-22 successfully held the nutrients in contact with the oil for the duration of the treatment period, despite constant exposure to the washing action of the surf and occasional heavy rainstorms. Today, the accelerator is routinely used in cleaning up all types of ordinary spills including diesel fuel spills along railway right-of-ways, truck yards and refinery sludge. Conditions under which the application of the accelerator is likely to be most successful are described

  17. 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.

  18. Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

    International Nuclear Information System (INIS)

    Dufay-Chanat, L; Bremer, J; Casas-Cubillos, J; Koettig, T; Vauthier, N; Van Weelderen, R; Winkler, T; Chorowski, M; Grabowski, M; Jedrusyna, A; Lindell, G; Nonis, M

    2015-01-01

    The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium were released under standard operational tunnel conditions. The data recorded include oxygen concentration, temperature and flow speed measurements, and video footage used to assess qualitatively the visibility. These measurements have been made in the up- and downstream directions, with respect to the air ventilation flow, of the spill point.This paper presents the experimental set-up under which these release tests were made, the effects of these releases on the atmospheric tunnel condition as a function of the release flow rate. We discuss the modification to the personnel access conditions to the LHC tunnel that are presently implemented as a result of these tests. (paper)

  19. Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

    Science.gov (United States)

    Dufay-Chanat, L.; Bremer, J.; Casas-Cubillos, J.; Chorowski, M.; Grabowski, M.; Jedrusyna, A.; Lindell, G.; Nonis, M.; Koettig, T.; Vauthier, N.; van Weelderen, R.; Winkler, T.

    2015-12-01

    The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium were released under standard operational tunnel conditions. The data recorded include oxygen concentration, temperature and flow speed measurements, and video footage used to assess qualitatively the visibility. These measurements have been made in the up- and downstream directions, with respect to the air ventilation flow, of the spill point. This paper presents the experimental set-up under which these release tests were made, the effects of these releases on the atmospheric tunnel condition as a function of the release flow rate. We discuss the modification to the personnel access conditions to the LHC tunnel that are presently implemented as a result of these tests.

  20. Application of EM holographic methods to borehole vertical electric source data to map a fuel oil spill

    International Nuclear Information System (INIS)

    Bartel, L.C.

    1993-01-01

    The multifrequency, multisource holographic method used in the analysis of seismic data is to extended electromagnetic (EM) data within the audio frequency range. The method is applied to the secondary magnetic fields produced by a borehole, vertical electric source (VES). The holographic method is a numerical reconstruction procedure based on the double focusing principle for both the source array and the receiver array. The approach used here is to Fourier transform the constructed image from frequency space to time space and set time equal to zero. The image is formed when the in-phase part (real part) is a maximum or the out-of-phase (imaginary part) is a minimum; i.e., the EM wave is phase coherent at its origination. In the application here the secondary magnetic fields are treated as scattered fields. In the numerical reconstruction, the seismic analog of the wave vector is used; i.e., the imaginary part of the actual wave vector is ignored. The multifrequency, multisource holographic method is applied to calculated model data and to actual field data acquired to map a diesel fuel oil spill

  1. Applications of EM holographic methods to borehole vertical electric source data to map a fuel oil spill

    International Nuclear Information System (INIS)

    Bartel, L.C.

    1993-01-01

    The multifrequency, multisource holographic method used in the analysis of seismic data is to extended electromagnetic (EM) data within the audio frequency range. The method is applied to the secondary magnetic fields produced by a borehole, vertical electric source (VES). The holographic method is a numerical reconstruction procedure based on the double focusing principle for both the source array and the receiver array. The approach used here is to Fourier transform the constructed image from frequency space to time space and set time equal to zero. The image is formed when the in-phase part (real part) is a maximum or the out-of-phase (imaginary part) is a minimum; i.e., the EM wave is phase coherent at its origination. In the application here the secondary magnetic fields are treated as scattered fields. In the numerical reconstruction, the seismic analog of the wave vector is used; i.e., the imaginary part of the actual wave vector is ignore. The multifrequency, multisource holographic method is applied to calculated model data and to actual field data acquired to map a diesel fuel oil spill

  2. 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

  3. Closure Report for Corrective Action Unit 330: Areas 6, 22, and 23 Tanks and Spill Sites, Nevada Test Site, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    A. T. Urbon

    2003-07-01

    This Closure Report (CR) documents the activities performed to close Corrective Action Unit (CAU) 330: Areas 6, 22, and 23 Tanks and Spill Sites, in accordance with the Federal Facility Agreement and Consent Order (FFACO of 1996), and the Nevada Division of Environmental Protection (NDEP)-approved Streamlined Approach for Environmental Restoration (SAFER) Plan for CAU 330: Areas 6, 22, and 23 Tanks and Spill Sites, Nevada Test Site (NTS), Nevada (U.S. Department of Energy, National Nuclear Security Administration Nevada Operation Office [NNSA/NV], 2001). CAU 330 consists of the following four Corrective Action Sites (CASs): 06-02-04, 22-99-06, 23-01-02, and 23-25-05 (Figure 1).

  4. 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%.

  5. Biodegradation of dispersed marine fuel oil in sediment under engineered pre-spill application strategy

    International Nuclear Information System (INIS)

    Hua, J.

    2006-01-01

    Biodegradation of marine fuel oil was studied by monitoring changes in residual oil and populations of microorganisms in marine sediments. Biodegradation rates for dispersant and soap water were 2.09 and 2.27 g/kg per day, respectively, under pre-application strategy, suggesting that the strategy may promote MFO dispersion and provide with sufficient source of food. The effect of temperature on the effectiveness of pre-application strategy is particularly obvious for the growth of fungi and Pseudomonas maltophilia. The effect of pre-application of soap water on the tolerance of aerobic bacteria, Escherichia coli, and P. maltophilia, was gradually diminished within 25-33 days. (author)

  6. In situ bioremediation of an underground diesel fuel spill: A case history

    Science.gov (United States)

    Frankenberger, W. T.; Emerson, K. D.; Turner, D. W.

    1989-05-01

    In the winter months of 1983, approximately 1000 gallons of diesel fuel had flowed along an asphalt parking lot of a commercial establishment towards a surface drain near an open creek. Investigations led to the discovery of an underground storage tank leaking diesel fuel. Exploratory borings showed that contamination was near the surface horizon and the capillary zone of the water table. Hydrocarbon quantities ranged up to 1500 mg/kg of soil. The plume continued to move in an eastward direction toward the surface water of the creek. A laboratory study indicated relatively high numbers of hydrocarbon-oxidizing organisms relative to glucose-utilizing microorganisms in the unsaturated vadose zone. Bioreclamation was initiated in April 1984 by injecting nutrients (nitrogen and phosphorus) and hydrogen peroxide and terminated in October 1984 upon no detection (<1 mg/kg) of hydrocarbons. A verification boring within the vicinity of the contaminated plume confirmed that residual contamination had attained background levels. The monitoring program was terminated in January 1987.

  7. 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.

  8. 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)

  9. Oil spills

    International Nuclear Information System (INIS)

    Spaulding, M.L.; Reed, M.

    1990-01-01

    Public awareness and concern for the potential short and long term impacts of oil spills on the marine environment has generally been high, particularly for regions of special ecological importance or where significant numbers of marine mammals and birds are present. This awareness was further heightened by the extraordinary number of recent large spills in coastal U.S. water: Exxon Valdez, Alaska; World Prodigy, Rhode Island; Presidente Rivera, Delaware; Rachel-B, Texas and American Trader, California. The occurrence of so many spills in a one year period is unprecedented in U.S. spill history. The legislative response to these spills has been immediate. New legislative initiative are rapidly being developed. Improved ways to organize spill response efforts are being devised and implemented. Funds are being allocated to further develop and improve spill response equipment and damage assessment methodologies. These spill events will have a significant impact in both the short and long term on oil exploration, development and transport in marine waters. They will result in major changes in management and operation of oil exploration and development. The purpose of this conference was to provide a forum for discussion of the changes which are currently taking place in oil spill legislation, management, and response strategies

  10. 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.

  11. 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)

  12. 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

  13. Monitored natural attenuation (MNA) and risk management applied to an active industrial site affected by fuel spill in groundwater

    International Nuclear Information System (INIS)

    De Pablo, J.; Marti, V.; Rovira, M.; Vinolas, C.; Navarro, O.

    2005-01-01

    Monitored Natural Attenuation (MNA) applied to sites were groundwater have been affected by a fuel spill from an Industrial Underground Storage Tank (UST) is economically viable and a reliable methodology to achieve remediation goals. MNA process consists in the control of naturally occurring physical, chemical , and biological processes and is based in the knowledge of the processes that take place and reduce the charge of compounds derived from fuel in the site of study. Because the risk for Human Health and Ecosystem define the concept of contaminant, during MNA special attention has to be taken on concentration diminution of that are or could become contaminants and in this way is possible to perform Risk-Based Land Management (RBLM) by measuring both, the primary lines of evidence (shrinking or stable plume of contaminants) and secondary lines of evidence (given by geochemical indicators in the plume). Once, evidences have been gathered, is possible to calculate the rate of attenuation of contaminants and evaluate if admissible risk is reached an in a reasonable time framework, in order to propose MNA as a unique remediation or combined with other procedures to apply to an affected site. The objective of the present study is to evaluate the application of MNA to an active industrial site in order to develop a RBLM able to assess that the risk for Human Health and ecosystem are acceptable. The added attractive of this methodology is the non-intrusiveness that allows not to stop the industrial activity. The site considered in our study is in an active company located about 15 Km to NW from Barcelona, Spain.The company has a buried UST containing heavy fuel oil for energetic use. During 2002 a general soil impact study revealed that subsoil and groundwater close to the UST were affected by hydrocarbon losses from the tank and in January 2003 the fuel of the tank was emptied by pumping. The free phase of fuel floating on groundwater remained on the aquifer. As a

  14. 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

  15. 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

  16. 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)

  17. 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)

  18. 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...

  19. 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.

  20. On site remediation of a fuel spill and soil reuse in Antarctica.

    Science.gov (United States)

    McWatters, R S; Wilkins, D; Spedding, T; Hince, G; Raymond, B; Lagerewskij, G; Terry, D; Wise, L; Snape, I

    2016-11-15

    The first large-scale remediation of fuel contamination in Antarctica treated 10000L of diesel dispersed in 1700t of soil, and demonstrated the efficacy of on-site bioremediation. The project progressed through initial site assessment and natural attenuation, passive groundwater management, then active remediation and the managed reuse of soil. Monitoring natural attenuation for the first 12years showed contaminant levels in surface soil remained elevated, averaging 5000mg/kg. By contrast, in five years of active remediation (excavation and biopile treatment) contaminant levels decreased by a factor of four. Chemical indicators showed hydrocarbon loss was apportioned to both biodegradation and evaporative processes. Hydrocarbon degradation rates were assessed against biopile soil temperatures, showing a phase of rapid degradation (first 100days above soil temperature threshold of 0°C) followed by slower degradation (beyond 100days above threshold). The biopiles operated successfully within constraints typical of harsh climates and remote sites, including limitations on resources, no external energy inputs and short field seasons. Non-native microorganisms (e.g. inoculations) and other organic materials (e.g. bulking agents) are prohibited in Antarctica making this cold region more challenging for remediation than the Arctic. Biopile operations included an initial fertiliser application, biannual mechanical turning of the soil and minimal leachate recirculation. The biopiles are a practical approach to remediate large quantities of contaminated soil in the Antarctic and already 370t have been reused in a building foundation. The findings presented demonstrate that bioremediation is a viable strategy for Antarctica and other cold regions. Operators can potentially use the modelled relationship between days above 0°C (threshold temperature) and the change in degradation rates to estimate how long it would take to remediate other sites using the biopile technology

  1. Oil Spills

    Science.gov (United States)

    ... up. How Oil Harms Animals and Plants in Marine Environments In general, oil spills can affect animals and plants in two ways: from the oil ... up. How Oil Harms Animals and Plants in Marine Environments In general, oil spills can affect animals and plants in two ways: from the oil ...

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. 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)

  9. 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

  10. 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

  11. 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...

  12. 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)

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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%

  18. 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)

  19. 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.)

  20. 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

  1. CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 390: AREAS 9, 10, AND 12 SPILL SITES, NEVADA TEST SITE, NEVADA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-10-01

    Corrective Action Unit (CAU) 390 consists four Corrective Action Sites (CASs) located in Areas 9, 10, and 12 of the Nevada Test Site. The closure activities performed at the CASs include: (1) CAS 09-99-03, Wax, Paraffin: 2 cubic yards of drilling polymer was removed on June 20,2005, and transported to the Area 9 Landfill for disposal. (2) CAS 10-99-01, Epoxy Tar Spill: 2 cubic feet of asphalt waste was removed on June 20,2005, and transported to the Area 9 Landfill for disposal. (3) CAS 10-99-03, Tar Spills: 3 cubic yards of deteriorated asphalt waste was removed on June 20,2005, and transported to the Area 9 Landfill for disposal. (4) CAS 12-25-03, Oil Stains (2); Container: Approximately 16 ounces of used oil were removed from ventilation equipment on June 28,2005, and recycled. One CAS 10-22-19, Drums, Stains, was originally part of CAU 390 but was transferred out of CAU 390 and into CAU 550, Drums, Batteries, and Lead Materials. The transfer was approved by the Nevada Division of Environmental Protection on August 19,2005, and a copy of the approval letter is included in Appendix D of this report.

  2. Improving oil classification quality from oil spill fingerprint beyond six sigma approach.

    Science.gov (United States)

    Juahir, Hafizan; Ismail, Azimah; Mohamed, Saiful Bahri; Toriman, Mohd Ekhwan; Kassim, Azlina Md; Zain, Sharifuddin Md; Ahmad, Wan Kamaruzaman Wan; Wah, Wong Kok; Zali, Munirah Abdul; Retnam, Ananthy; Taib, Mohd Zaki Mohd; Mokhtar, Mazlin

    2017-07-15

    This study involves the use of quality engineering in oil spill classification based on oil spill fingerprinting from GC-FID and GC-MS employing the six-sigma approach. The oil spills are recovered from various water areas of Peninsular Malaysia and Sabah (East Malaysia). The study approach used six sigma methodologies that effectively serve as the problem solving in oil classification extracted from the complex mixtures of oil spilled dataset. The analysis of six sigma link with the quality engineering improved the organizational performance to achieve its objectivity of the environmental forensics. The study reveals that oil spills are discriminated into four groups' viz. diesel, hydrocarbon fuel oil (HFO), mixture oil lubricant and fuel oil (MOLFO) and waste oil (WO) according to the similarity of the intrinsic chemical properties. Through the validation, it confirmed that four discriminant component, diesel, hydrocarbon fuel oil (HFO), mixture oil lubricant and fuel oil (MOLFO) and waste oil (WO) dominate the oil types with a total variance of 99.51% with ANOVA giving F stat >F critical at 95% confidence level and a Chi Square goodness test of 74.87. Results obtained from this study reveals that by employing six-sigma approach in a data-driven problem such as in the case of oil spill classification, good decision making can be expedited. Copyright © 2017. Published by Elsevier Ltd.

  3. 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)

  4. Weathering of hydrocarbons in mangrove sediments: testing the effects of using dispersants to treat oil spills

    International Nuclear Information System (INIS)

    Burns, K.A.; Codi, S.; Pratt, C.; Duke, N.C.

    1999-01-01

    This field study was a combined chemical and biological investigation of the relative effects of using dispersants to treat oil spills impacting mangrove habitats. The aim of the chemistry was to determine whether dispersant affected the short- or long-term composition of a medium range crude oil (Gippsland) stranded in a tropical mangrove environment in Queensland, Australia. Sediment cores from three replicate plots of each treatment (oil only and oil plus dispersant) were analyzed for total hydrocarbons and for individual molecular markers (alkanes, aromatics, triterpanes, and steranes). Sediments were collected at 2 days, then 1, 7, 13 and 22 months post-spill. Over this time, oil in the six treated plots decreased exponentially from 36.6 ± 16.5 to 1.2 ± 0.8 mg/g dry wt. There was no statistical difference in initial oil concentrations, penetration of oil to depth, or in the rates of oil dissipation between oiled or dispersed oil plots. At 13 months, alkanes were > 50% degraded, aromatics were ∼30% degraded based upon ratios of labile to resistant markers. However, there was no change in the triterpane or sterane biomarker signatures of the retained oil. This is of general forensic interest for pollution events. The predominant removal processes were evaporation (≤27%) and dissolution (≥56%), with a lag-phase of 1 month before the start of significant microbial degradation (≤17%). The most resistant fraction of the oil that remained after 7 months (the higher molecular weight hydrocarbons) correlated with the initial total organic carbon content of the soil. Removal rate in the Queensland mangroves was significantly faster than that observed in the Caribbean and was related to tidal flushing. (author)

  5. 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

  6. FUELS IN SOIL TEST KIT: FIELD USE OF DIESEL DOG SOIL TEST KITS

    Energy Technology Data Exchange (ETDEWEB)

    Susan S. Sorini; John F. Schabron; Joseph F. Rovani, Jr.

    2002-09-30

    Western Research Institute (WRI) has developed a new commercial product ready for technology transfer, the Diesel Dog{reg_sign} Portable Soil Test Kit, for performing analysis of fuel-contaminated soils in the field. The technology consists of a method developed by WRI (U.S. Patents 5,561,065 and 5,976,883) and hardware developed by WRI that allows the method to be performed in the field (patent pending). The method is very simple and does not require the use of highly toxic reagents. The aromatic components in a soil extract are measured by absorption at 254 nm with a field-portable photometer. WRI added significant value to the technology by taking the method through the American Society for Testing and Materials (ASTM) approval and validation processes. The method is designated as ASTM Method D 5831-96, Standard Test Method for Screening Fuels in Soils. This ASTM designation allows the method to be used for federal compliance activities. In June 2001, the Diesel Dog technology won an American Chemical Society Regional Industrial Innovations Award. To gain field experience with the new technology, Diesel Dog kits have been used for a variety of site evaluation and cleanup activities. Information gained from these activities has led to improvements in hardware configurations and additional insight into correlating Diesel Dog results with results from laboratory methods. The Wyoming Department of Environmental Quality (DEQ) used Diesel Dog Soil Test Kits to guide cleanups at a variety of sites throughout the state. ENSR, of Acton, Massachusetts, used a Diesel Dog Portable Soil Test Kit to evaluate sites in the Virgin Islands and Georgia. ChemTrack and the U.S. Army Corps of Engineers successfully used a test kit to guide excavation at an abandoned FAA fuel-contaminated site near Fairbanks, Alaska. Barenco, Inc. is using a Diesel Dog Portable Soil Test Kit for site evaluations in Canada. A small spill of diesel fuel was cleaned up in Laramie, Wyoming using a Diesel

  7. 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....

  8. Worldwide analysis of marine oil spill cleanup cost factors

    International Nuclear Information System (INIS)

    Etkin, D.S.

    2000-01-01

    The many factors that influence oil spill response costs were discussed with particular emphasis on how spill responses differ around the world because of differing cultural values, socio-economic factors and labor costs. This paper presented an analysis of marine oil spill cleanup costs based on the country, proximity to shoreline, spill size, oil type, degree of shoreline oiling and cleanup methodology. The objective was to determine how each factor impacts per-unit cleanup costs. Near-shore spills and in-port spills were found to be 4-5 times more expensive to clean than offshore spills. Responses to spills of heavy fuels also cost 10 times more than for lighter crudes and diesel. Spill responses for spills under 30 tonnes are 10 times more costly than on a per-unit basis, for spills of 300 tonnes. A newly developed modelling technique that can be used on different types of marine spills was described. It is based on updated cost data acquired from case studies of more than 300 spills in 40 countries. The model determines a per-unit cleanup cost estimation by taking into consideration oil type, location, spill size, cleanup methodology, and shoreline oiling. It was concluded that the actual spill costs are totally dependent on the actual circumstances of the spill. 13 refs., 10 tabs., 3 figs

  9. 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 ...

  10. Investigation of the presence of toxic components of petroleum hydrocarbons in Guanabara Bay, Brazil following the 2000 PETROBRAS fuel oil spill

    Energy Technology Data Exchange (ETDEWEB)

    Romao, Catia Maria [Instituto Brasileiro do Meio Ambiente e dos Recursos Naturais Renovaveis (IBAMA), Rio de Janeiro, RJ (Brazil). Escritorio de Licenciamento de Petroleo e Nuclear; Vleet, Edward S. Van

    2003-07-01

    On January 18, 2000, approximately 340,000 gallons of marine fuel 380 oil were released into Guanabara Bay, Rio de Janeiro, Brazil, as a consequence of a pipeline transfer accident at the Duque de Caxias Refinery (PETROBRAS). Two years after the spill, the present investigation (sponsored by Center for Disaster Management and Humanitarian Assistance - College of Public Health - University of South Florida) was conducted to assess the levels of Polycyclic Aromatic Hydrocarbons (PAHs) on samples of water, sediments and edible tissue of the fishes (Mullet - Mugilliza and Croaker - Micropogonias furnieri) collected using two types of device (nets and fish traps) from the spill area in July and August 2002. The fishes samples collected in both months were considered to range from being not contaminated to being moderately contaminated by PAHs. Among all the sediments, only one (Point 10, July 2002) showed a total PAH concentration representing highly contaminated conditions. Except for Point 10, all other sediments could be considered minimally to moderately contaminated. Dissolved PAH concentrations found in the water samples were considered to range from minimally to moderately contaminated. (author)

  11. 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...

  12. 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...

  13. 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

  14. A local oil spill revisited

    International Nuclear Information System (INIS)

    Teal, J.M.

    1993-01-01

    In October 1969 George Hampson and Howard Sanders (Woods Hole Oceanographic Institution) described a 'Local Oil Spill' in Oceanus. The spill had occurred a month before when the barge Florida, loaded with no. 2 fuel oil, ran into some rocks in Buzzards Bay off West Falmouth, Massachusetts. In the summer of 1989, almost 20 years later, They visited the Wild Harbor marsh area that had suffered the greatest impact from the spill to see if any traces of the event in the marsh ecosystem could be found. During those 20 years, the site has been visited by graduate students in marine ecology, by reporters seeking information about current oil spills but also interested in seeing the effects of the Wild Harbor spill, and by visiting scientists curious about one of the world's best-studied oil spills. For more than a decade after the spill, an oil sheen appeared on the surface of the water when mud from the most heavily oiled parts of the marsh was disturbed. During the second decade, the marsh's appearance returned to normal

  15. 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)

  16. 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.

  17. 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)

  18. 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)

  19. 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

  20. 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.

  1. 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...

  2. 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...

  3. Introduction of CFD Analysis for Consideration of Fuel Ignition Test

    OpenAIRE

    岡崎,航介; 吉田,肇; 入澤,優麿; 櫻庭,隆貴

    2014-01-01

    Not a few fires in the engine room of ships are caused from the ignition of spilled oil on the hot part of machines such as turbo-chargers, boilers etc. Authors have continued to investigate ignition and combustion phenomenon of oils on the hot surface of a metal plate with various conditions of the surface and many kinds of fuels for understanding of the mechanism of fires in engine rooms. However, it could not be well-considered about flow of the air in the reaction region because of the de...

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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)

  10. Spatial data quality and coastal spill modelling

    International Nuclear Information System (INIS)

    Li, Y.; Brimicombe, A.J.; Ralphs, M.P.

    1998-01-01

    Issues of spatial data quality are central to the whole oil spill modelling process. Both model and data quality performance issues should be considered as indispensable parts of a complete oil spill model specification and testing procedure. This paper presents initial results of research that will emphasise to modeler and manager alike the practical issues of spatial data quality for coastal oil spill modelling. It is centred around a case study of Jiao Zhou Bay in the People's Republic of China. The implications for coastal oil spill modelling are discussed and some strategies for managing the effects of spatial data quality in the outputs of oil spill modelling are explored. (author)

  11. 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

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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...

  2. 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)

  3. 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

  4. Corrective Action Investigation Plan for Corrective Action Unit 234: Mud Pits, Cellars, and Mud Spills, Nevada Test Site, Nevada, Revision 0

    International Nuclear Information System (INIS)

    Grant Evenson

    2007-01-01

    Corrective Action Unit 234, Mud Pits, Cellars, and Mud Spills, consists of 12 inactive sites located in the north and northeast section of the NTS. The 12 CAU 234 sites consist of mud pits, mud spills, mud sumps, and an open post-test cellar. The CAU 234 sites were all used to support nuclear testing conducted in the Yucca Flat and Rainier Mesa areas during the 1950s through the 1970s. The CASs in CAU 234 are being investigated because hazardous and/or radioactive constituents may be present in concentrations that could potentially pose a threat to human health and the environment. Existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives for the CASs. Additional information will be generated by conducting a CAI before evaluating and selecting appropriate corrective action alternatives

  5. 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...

  6. Source identification of underground fuel spills in a petroleum refinery using fingerprinting techniques and chemo-metric analysis. A Case Study

    International Nuclear Information System (INIS)

    Kanellopoulou, G.; Gidarakos, E.; Pasadakis, N.

    2005-01-01

    Crude oil and its refining products are the most frequent contaminants, found in the environment due to spills. The aim of this work was the identification of spill source(s) in the subsurface of a petroleum refinery. Free phase samples were analyzed with gas chromatography and the analytical results were interpreted using Principal Component Analysis (PCA) method. The chemical analysis of groundwater samples from the refinery subsurface was also employed to obtain a comprehensive picture of the spill distribution and origin. (authors)

  7. 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)

  8. 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

  9. 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...

  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. 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

  12. 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.

  13. 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.)

  14. 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

  15. 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

  16. 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

  17. 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.

  18. 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...

  19. 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

  20. 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.).

  1. Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

    CERN Document Server

    Dufay-Chanat, L; Casas-Cubillos, J; Chorowski, M; Grabowski, M; Jedrusyna, A; Lindell, G; Nonis, M; Koettig, T; Vauthier, N; van Weelderen, R; Winkler, T

    2015-01-01

    The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium wer...

  2. 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

  3. 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.

  4. 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

  5. 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

  6. Analysis of the October 5, 1979 lithium spill and fire in the Lithium Processing Test Loop

    International Nuclear Information System (INIS)

    Maroni, V.A.; Beatty, R.A.; Brown, H.L.; Coleman, L.F.; Foose, R.M.; McPheeters, C.C.; Slawecki, M.; Smith, D.L.; Van Deventer, E.H.; Weston, J.R.

    1981-12-01

    On October 5, 1979, the Lithium Processing Test Loop (LPTL) developed a lithium leak in the electromagnetic (EM) pump channel, which damaged the pump, its surrounding support structure, and the underlying floor pan. A thorough analysis of the causes and consequences of the pump failure was conducted by personnel from CEN and several other ANL divisions. Metallurgical analyses of the elliptical pump channel and adjacent piping revealed that there was a significant buildup of iron-rich crystallites and other solid material in the region of the current-carrying bus bars (region of high magnetic field), which may have resulted in a flow restriction that contributed to the deterioration of the channel walls. The location of the failure was in a region of high residual stress (due to cold work produced during channel fabrication); this failure is typical of other cold work/stress-related failures encountered in components operated in forced-circulation lithium loops. Another important result was the isolation of crystals of a compound characterized as Li/sub x/CrN/sub y/. Compounds of this type are believed to be responsible for much of the Fe, Cr, and Ni mass transfer encountered in lithium loops constructed of stainless steel. The importance of nitrogen in the mass-transfer mechanism has long been suspected, but the existence of stable ternary Li-M-N compounds (M = Fe, Cr, Ni) had not previously been verified

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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.)

  19. 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

  20. 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.

  1. 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.

  2. 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...

  3. 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)

  4. 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.

  5. 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.)

  6. 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)

  7. 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.

  8. 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

  9. 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)

  10. 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)

  11. 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)

  12. 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.

  13. 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

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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.)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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...

  7. 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

  8. The Worldwide Oil Spill Model (WOSM)

    International Nuclear Information System (INIS)

    Anderson, E.L.; Howlett, E.; Jayko, K.; Reed, M.; Spaulding, M.; Kolluru, V.

    1993-01-01

    The Worldwide Oil Spill Model (WOSM) is a standalone microcomputer-based state-of-the-art oil spill model system for use in oil spill response decision support, planning, research, training, and contingency planning. WOSM was developed under support provided by a consortium of oil companies and government agencies. WOSM represents the next generation of oil spill model beyond the OILMAP modelling system (Spaulding et al, 1992). WOSM is designed in a shell architecture in which the only parameters that change are those that describe the area in which the spill model is to be applied. A limited function geographic information system (GIS) is integrated within the model system, and the spill modelling shell has been extended to include interfaces to other GIS systems and digital data. WOSM contains all the databases, data manipulation and graphical display tools, and models to simulate any type of oil spill. The user has control over which weathering processes are to be modelled, and WOSM data input tools enable continual refinement of model predictions as more refined data is imported. Use of WOSM is described and illustrated, showing sample screens and applications. WOSM algorithms and file structure are also outlined. An example test case of a spill in the Juan de Fuca strait is included. 29 refs., 7 figs., 1 tab

  9. 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)

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. Evolution of paraffinic and naphtenic hydrocarbon and 3,4 benzopyrene content in mussels from a coastal zone polluted by a fuel spill

    Energy Technology Data Exchange (ETDEWEB)

    Bories, G; Tulliez, J; Peltier, J C; Fleckinger, R

    1969-05-03

    After an oil spill, a coastal zone was polluted and wild mussels were contaminated by paraffinic and naphtenic hydrocarbons and 3,4-benzopyrene. The evolution of this contamination was followed. Normal levels were re-established after a month and a half. Normal paraffins were metabolized faster than other hydrocarbons.

  9. 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)

  10. Identification of spilled oils by NIR spectroscopy technology based on KPCA and LSSVM

    Science.gov (United States)

    Tan, Ailing; Bi, Weihong

    2011-08-01

    Oil spills on the sea surface are seen relatively often with the development of the petroleum exploitation and transportation of the sea. Oil spills are great threat to the marine environment and the ecosystem, thus the oil pollution in the ocean becomes an urgent topic in the environmental protection. To develop the oil spill accident treatment program and track the source of the spilled oils, a novel qualitative identification method combined Kernel Principal Component Analysis (KPCA) and Least Square Support Vector Machine (LSSVM) was proposed. The proposed method adapt Fourier transform NIR spectrophotometer to collect the NIR spectral data of simulated gasoline, diesel fuel and kerosene oil spills samples and do some pretreatments to the original spectrum. We use the KPCA algorithm which is an extension of Principal Component Analysis (PCA) using techniques of kernel methods to extract nonlinear features of the preprocessed spectrum. Support Vector Machines (SVM) is a powerful methodology for solving spectral classification tasks in chemometrics. LSSVM are reformulations to the standard SVMs which lead to solving a system of linear equations. So a LSSVM multiclass classification model was designed which using Error Correcting Output Code (ECOC) method borrowing the idea of error correcting codes used for correcting bit errors in transmission channels. The most common and reliable approach to parameter selection is to decide on parameter ranges, and to then do a grid search over the parameter space to find the optimal model parameters. To test the proposed method, 375 spilled oil samples of unknown type were selected to study. The optimal model has the best identification capabilities with the accuracy of 97.8%. Experimental results show that the proposed KPCA plus LSSVM qualitative analysis method of near infrared spectroscopy has good recognition result, which could work as a new method for rapid identification of spilled oils.

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. Spread and burning behavior of continuous spill fires

    DEFF Research Database (Denmark)

    Zhao, Jinlong; Huang, Hong; Jomaas, Grunde

    2017-01-01

    Spill fire experiments with continuous discharge on a fireproof glass sheet were conducted to improve the understanding of spill fire spread and burning. Ethanol was used as the fuel and the discharge rate was varied from 2.8. mL/s to 7.6. mL/s. Three ignition conditions were used...... in the experiments; no ignition, instantaneous ignition and delayed ignition. The spread rate, regression rate, penetrated thermal radiation and the temperature of the bottom glass were analyzed. The experiments clearly show the entire spread process for spill fires. Further, the regression rate of spill fires...... at the quasi-steady burning was lower than that of pool fires and the ratio of the spill fires' regression rate to the pool fires' regression rate was found to be approximately 0.89. With respect to the radiative penetration and the heat conduction between the fuel layer and the glass, a regression rate...

  2. 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.

  3. Progress toward the development of micro- and meso-scale methods for predicting the behavior of low-API gravity oils (LAPIO) spilled on water

    International Nuclear Information System (INIS)

    Ostazeski, S.A.; Durell, G.S.; Uhler, A.D.

    1996-01-01

    Bench-scale weathering studies were conducted on a number of low-API gravity oils (LAPIO), and flume tank studies were conducted on several LAPIO products. The objective was to develop a method which would predict the environmental behaviour of LAPIO fuel spilled on water. Unlike crude oils and petroleum products which float when spilled on water, LAPIO spills are unpredictable. LAPIO fuels, which are used for electric power generation, have densities greater or equal to 1.0 g/ml. When spilled at sea, they may float, be neutrally buoyant, or sink, depending on the conditions of the receiving water. Their behavior depends on the physical and chemical properties of the specific oil and also the temperature and composition of the water. In order to simulate the behavior of oil spilled at sea, a number of different LAPIO fuels were artificially weathered by evaporative distillation. Tests were conducted to determine the physical and chemical properties of the oil residues, the buoyancy of the oils, their emulsifying properties and chemical dispersability. Results have shown that current laboratory methods for determining emulsion formation, stability, and kinetics are not adequate for these highly viscous fuels. Laboratory methods need refining and improving. 7 refs

  4. 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)

  5. 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)

  6. 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.

  7. 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)

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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)

  13. 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

  14. 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)

  15. 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.

  16. 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.

  17. 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

  18. Novel apparatus permitting the recovery of polluting products such as hydrocarbons and fuel oil spilled onto the surface of the water

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-26

    This apparatus for the recovery of oil spills consists of a chassis equipped at both ends with drums rotating and carrying a closed loop of a conveyor belt made of a sponge-like or foam-like absorbing substance which resists the effect of hydrocarbons, oils, oxidants and alkaline materials, while its water absorption power is practically zero. This material is glued or cast on a fabric core or on a large mesh material. A cleaning drum associated with the belt compresses the belt and causes the polluting material to be squeezed into a vat. The chassis is mounted on a ship with shallow draft or on an amphibious vehicle. This apparatus is applicable to all cases where a polluting product spilled onto the surface of the water must be eliminated.

  19. 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)

  20. 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

  1. 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.

  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. 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)

  7. 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.

  8. 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

  9. Chemometric techniques in oil classification from oil spill fingerprinting.

    Science.gov (United States)

    Ismail, Azimah; Toriman, Mohd Ekhwan; Juahir, Hafizan; Kassim, Azlina Md; Zain, Sharifuddin Md; Ahmad, Wan Kamaruzaman Wan; Wong, Kok Fah; Retnam, Ananthy; Zali, Munirah Abdul; Mokhtar, Mazlin; Yusri, Mohd Ayub

    2016-10-15

    Extended use of GC-FID and GC-MS in oil spill fingerprinting and matching is significantly important for oil classification from the oil spill sources collected from various areas of Peninsular Malaysia and Sabah (East Malaysia). Oil spill fingerprinting from GC-FID and GC-MS coupled with chemometric techniques (discriminant analysis and principal component analysis) is used as a diagnostic tool to classify the types of oil polluting the water. Clustering and discrimination of oil spill compounds in the water from the actual site of oil spill events are divided into four groups viz. diesel, Heavy Fuel Oil (HFO), Mixture Oil containing Light Fuel Oil (MOLFO) and Waste Oil (WO) according to the similarity of their intrinsic chemical properties. Principal component analysis (PCA) demonstrates that diesel, HFO, MOLFO and WO are types of oil or oil products from complex oil mixtures with a total variance of 85.34% and are identified with various anthropogenic activities related to either intentional releasing of oil or accidental discharge of oil into the environment. Our results show that the use of chemometric techniques is significant in providing independent validation for classifying the types of spilled oil in the investigation of oil spill pollution in Malaysia. This, in consequence would result in cost and time saving in identification of the oil spill sources. Copyright © 2016. Published by Elsevier Ltd.

  10. Enhancing spill prevention and response preparedness through quality control techniques

    International Nuclear Information System (INIS)

    Jones, M.A.; Butts, R.L.; Pickering, T.H.; Lindsay, J.R.; McCully, B.S.

    1993-01-01

    The year 1990 saw passage of federal and state oil spill legislation directing the US Environmental Protection Agency and the Florida Department of Environmental Regulation to require on shore bulk petroleum storage facilities to improve their oil spill response and prevention capabilities. The Florida Power ampersand Light Company (FPL), to address concerns arising out of several recent significant spills which had occurred worldwide, and to examine its current situation with regard compliance with the new laws, formed a quality improvement interdepartmental task team in July 1989. Its mission was to reduce the potential for oil spills during waterborne transportation between FPL's fuel oil terminals and its power plants and during transfer and storage of oil at these facilities. Another objective of the team was to enhance the company's spill response preparedness. Using quality control tools and reliability techniques, the team conducted a detailed analysis of seven coastal power plants and five fuel oil terminal facilities. This analysis began with the development of cause-and-effect diagrams designed to identify the root causes of spills so that corrective and preventive actions could be taken. These diagram are constructed by listing possible causes of oil spills under various major categories of possible system breakdown, such as man, method, equipment, and materials. Next, potential root causes are identified and then verified. The team identified the occurrence of surface water oil spill and reduced spill response capability as primary concerns and accordingly constructed cause-and-effect diagrams for both components. Lack of proper procedures, failure of control equipment, and inadequate facility design were identified as potential root causes leading to surface water oil spills. Lack of proper procedures, an inconsistent training program, and response equipment limitations were identified as potential root causes affecting oil spill response capabilities

  11. 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

  12. 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

  13. 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)

  14. 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

  15. Oil Spill Response Manual

    NARCIS (Netherlands)

    Marieke Zeinstra; Sandra Heins; Wierd Koops

    2014-01-01

    A two year programme has been carried out by the NHL University of Applied Sciences together with private companies in the field of oil and chemical spill response to finalize these manuals on oil and chemical spill response. These manuals give a good overview of all aspects of oil and chemical

  16. Gas spill emergency

    International Nuclear Information System (INIS)

    1997-01-01

    This video presentation was designed to explain the steps that should be taken in the event of a petroleum product spill on land, to keep damages and consequences to a minimum. The events that took place when an oil truck full of gasoline overturned and smashed into a house on a residential street were described to illustrate the principles involved. The following sequence of events and actions, based on general principles of bringing the situation under control during an emergency operation were depicted: (1) identification of spilled product, (2) assessment of the situation, (3) setting priorities and evacuating the endangered area, and (4) setting up a communication system. The fire fighters sprayed the area with foam because of the fire and explosion potential. Sand was used to contain the spill and to keep it out of the storm sewers. The spilled oil was recovered. Three other spill situations - a spill at a service station, a spill in a ditch, and a spill in a waterway - were also documented. It was emphasized that while it is not possible to establish a single set of rules and actions that would apply to all situations since no two accidents involving petroleum products are alike, the general principles are universal and can be applied in all situations. First priority to consider should always be human life, then property, then the environment

  17. 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.)

  18. 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

  19. 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

  20. 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.

  1. Recent spill experiences: the North Cape spill

    International Nuclear Information System (INIS)

    Spaulding, M.

    1996-01-01

    The events which surrounded the spill of some 825,000 gallons of heating oil into the sea off the Rhode Island coast in January 1996, were described. Birds and marine life (lobsters in particular) had been affected by the spill, with injured and dead animals appearing from first light. Wildlife handling procedures were established immediately. A significant amount of the spilled oil had entered into the water column and could not be seen or controlled, and was moving along the coast into coastal ponds. Considerable seafood contamination was inevitable. To avoid even greater problems, it was decided to close the south shore beaches to the public and 105 square miles of coastal area to fishing and shell fishing. 4 refs., 2 figs

  2. Recent spill experiences: the North Cape spill

    Energy Technology Data Exchange (ETDEWEB)

    Spaulding, M. [Rhode Island Univ., Kingston, RI (United States). Dept. of Ocean Engineering

    1996-09-01

    The events which surrounded the spill of some 825,000 gallons of heating oil into the sea off the Rhode Island coast in January 1996, were described. Birds and marine life (lobsters in particular) had been affected by the spill, with injured and dead animals appearing from first light. Wildlife handling procedures were established immediately. A significant amount of the spilled oil had entered into the water column and could not be seen or controlled, and was moving along the coast into coastal ponds. Considerable seafood contamination was inevitable. To avoid even greater problems, it was decided to close the south shore beaches to the public and 105 square miles of coastal area to fishing and shell fishing. 4 refs., 2 figs.

  3. Oil spill statistics and oil spill monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Viebahn, C. von [Greifswald Univ. (Germany). Dept. of Geography

    2001-09-01

    The main parts of the report describe the analysis and it's results of German and international oil spill data (North Sea and Baltic Sea). In order to improve the current oil spill monitoring of the Baltic Sea regarding oil spill data, the report proposes the establishment of a combined monitoring system; its suitability is shown on selected examples. This contains today's pollution control aircraft plus in-service aircraft and satellites. (orig.) [German] Der Schwerpunkt der Arbeit liegt in der Analyse von Daten ueber marine Oelschadensfaelle in deutschen und internationalen Gewaessern (Nord- und Ostsee). Um die heutige Ueberwachung der Ostsee im Hinblick auf Oelschadensfaelle zu verbessern, wird die Einrichtung eines kombinierten Ueberwachungssystems vorgeschlagen und dessen Eignung an ausgewaehlten Beispielen dargestellt. Dieses umfasst sowohl die heute eingesetzten Ueberwachungsflugzeuge sowie zusaetzlich Linienflugzeuge und Satelliten. (orig.)

  4. 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

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. Responding to the Sea Empress oil spill

    International Nuclear Information System (INIS)

    Leonard, D.R.P.; Law, R.J.; Kelly, C.A.

    1999-01-01

    The Ministry of Agriculture, Fisheries and Food (MAFF) is a government department which has responsibility in England and in Wales (acting on behalf of the Secretary of State for Wales) for controlling deposits in the sea, including approving the use of dispersants in oil spill response. MAFF also has responsibility in relation to the management of sustainable commercial fish and shellfish fisheries. Following the grounding of the tanker Sea Empress on 15 February 1996, over 72,000 tonnes of crude oil and bunker fuel was lost. This paper summarises the involvement of MAFF staff in the response phase, and in the subsequent assessment of the environmental impact of the oil spill and the associated clean up operations on commercial fisheries. After two and a half years of environmental monitoring and complementary research, it is concluded that the oil spill has had an insignificant impact on these fisheries beyond their closure during the incident response phase. Suggestions for further work are discussed. (author)

  11. The Prestige oil spill response in the French coastal waters : setting up a second-line response system

    International Nuclear Information System (INIS)

    Kerambrun, L.; Lavenant, M.; Cariou, G.; Poisson, H.; Goasguen, H.; Peltier, M.

    2005-01-01

    A heavy fuel oil spill from the Prestige oil tanker in early 2003 threatened the coastline of France, particularly the sensitive sandy dunes of the Aquitaine coast. The spill posed a long-term threat to two affected tourist islands that host a rich marine life, including the Marennes-Oleron oysters and juvenile eel fisheries. An oil spill response system was established by the French Maritime Affairs Department. The two-tiered response system was constructed with dynamic and static components and was adapted for strong tidal currents and muddy waters where fuel patches from the spill might not be visible. Local fishermen with knowledge of the waters used their trawlers and eel-fishing boats that were equipped with oil recovery trawls. A local netting device was also tested. The second response line for shallow waters consisted of small boats with net bags. The storage, transfer and treatment of the oily waste were controlled and available oil waste treatment procedures were assessed at local, regional and national levels for each type of oil waste fuel. 2 refs

  12. 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

  13. 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

  14. 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

  15. 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...

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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...

  3. 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)

  4. 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

  5. Maximum tech, minimum time. Response and cleanup of the Fidalgo Bay oil spill

    International Nuclear Information System (INIS)

    Pintler, L.R.

    1991-01-01

    A booster pump failure on a pipeline at Texaco's Anacortes refinery spilled more than 17,000 gallons of oil into Fidalgo Bay. A description is given of the spill control measures taken under Texaco's Spill Prevention and Control Countermeasures and facility contingency plans. The spill was addressed quickly, and containment booms were used to cordon off the spill. Vacuum trucks, rope mop machines and disk skimmers were used to collect the thickest concentrations of oil, and the oil and water collected was separated at the refinery's wastewater treatment centre. Nonwoven polypropylene sorbent pads, sweeps, booms and oil snares were used to clean up thinner concentrations of oil. Essential steps for a smooth spill response include the following: a comprehensive spill prevention and control countermeasures plan, training and regular drills and testing; immediate notification of appropriate regulatory agencies and company emergency response personnel; and the use of professional oil spill management contractors to assist in spill cleanup. 2 figs

  6. 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)

  7. Pleural spill malign

    International Nuclear Information System (INIS)

    Camacho Duran, Fidel; Zamarriego, Roman; Gonzalez, Mauricio

    2002-01-01

    The pleural spills are developed because of an alteration in the mechanisms that usually move between 5 and 10 liters of liquid through the space pleural every 24 hours and this is reabsorbed, only leaving 5 to 20 ml present. The causes more common of spill pleural they are: congestive heart failure, bacterial pneumonia, malign neoplasia and pulmonary clot. The causes more common of pleural spill malign in general are: cancer of the lung, cancer of the breast and lymphomas. In the man, cancer of the lung, lymphomas and gastrointestinal cancer. In the woman, cancer of the breast, gynecological cancer and lung cancer. The paper, includes their characteristics, treatments and medicines

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. Trends in oil spills from tanker ships 1995-2004

    International Nuclear Information System (INIS)

    Huijer, K.

    2005-01-01

    The trends in oil spills around the world over from 1995 to 2004 were examined and analyzed for possible influences on spill volumes and frequencies for incidents of 3 spill size classes. The International Tanker Owners Pollution Federation (ITOPF) has maintained a database since 1974 of all oil spills from tankers, combined carriers and barges. The number of oil spills has decreased significantly in the last 30 years despite a steady increase in maritime oil trade since the 1980s. The recent trends were identified by causes, locations, oil type, and shipping legislation. The causes include ship loading/discharging, bunkering, collisions, groundings, hull failures and fires. The types of oil spilt include bunker, crude, cargo fuel, white product and some unknowns. It was concluded that the decline in oil spills is due to a range of initiatives taken by governments and the shipping industry rather than any one factor. Some notable influences towards reduced number of spills include: the international convention for the prevention of pollution from ships of 1972, as modified by the Protocol of 1978; the international convention for the safety of life at sea of 1974; and the Oil Pollution Act of 1990. Results of investigations into the causes of spills serve the purpose of informing the international process to further prevent and reduce marine oil pollution due to tankers. 7 refs., 5 tabs., 12 figs

  13. Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 544: Cellars, Mud Pits, and Oil Spills, Nevada Test Site, Nevada, Revision 0

    Energy Technology Data Exchange (ETDEWEB)

    Mark Krauss

    2010-07-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the actions needed to achieve closure for Corrective Action Unit (CAU) 544, Cellars, Mud Pits, and Oil Spills, identified in the Federal Facility Agreement and Consent Order (FFACO). Corrective Action Unit 544 comprises the following 20 corrective action sites (CASs) located in Areas 2, 7, 9, 10, 12, 19, and 20 of the Nevada Test Site (NTS): • 02-37-08, Cellar & Mud Pit • 02-37-09, Cellar & Mud Pit • 07-09-01, Mud Pit • 09-09-46, U-9itsx20 PS #1A Mud Pit • 10-09-01, Mud Pit • 12-09-03, Mud Pit • 19-09-01, Mud Pits (2) • 19-09-03, Mud Pit • 19-09-04, Mud Pit • 19-25-01, Oil Spill • 19-99-06, Waste Spill • 20-09-01, Mud Pits (2) • 20-09-02, Mud Pit • 20-09-03, Mud Pit • 20-09-04, Mud Pits (2) • 20-09-06, Mud Pit • 20-09-07, Mud Pit • 20-09-10, Mud Pit • 20-25-04, Oil Spills • 20-25-05, Oil Spills This plan provides the methodology for field activities needed to gather the necessary information for closing each CAS. There is sufficient information and process knowledge from historical documentation and investigations of similar sites regarding the expected nature and extent of potential contaminants to recommend closure of CAU 544 using the SAFER process. Using the approach approved for previous mud pit investigations (CAUs 530–535), 14 mud pits have been identified that • are either a single mud pit or a system of mud pits, • are not located in a radiologically posted area, and • have no evident biasing factors based on visual inspections. These 14 mud pits are recommended for no further action (NFA), and further field investigations will not be conducted. For the sites that do not meet the previously approved closure criteria, additional information will be obtained by conducting a field investigation before selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible

  14. 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.

  15. 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

  16. 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.

  17. 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

  18. 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...

  19. 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...

  20. 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

  1. 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

  2. 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.

  3. 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.

  4. Tunnel nitrogen spill experiment

    International Nuclear Information System (INIS)

    Ageyev, A.I.; Alferov, V.N.; Mulholland, G.T.

    1983-01-01

    The Energy Saver Safety Analysis Report (SAR) found the tunnel oxygen deficiency considerations emphasized helium spills. These reports concluded the helium quickly warms and because of its low denisty, rises to the apex of the tunnel. The oxygen content below the apex and in all but the immediate vicinity of the helium spill is essentially unchanged and guarantees an undisturbed source of oxygen especially important to fallen personnel. In contrast nitrogen spills warm slower than helium due to the ratio of the enthalpy changes per unit volume spilled spread more uniformly across the tunnel cross-section when warmed because of the much smaller density difference with air, and generally provides a greater hazard than helium spills as a result. In particular there was concern that personnel that might fall to the floor for oxygen deficiency or other reasons might find less, and not more, oxygen with dire consequences. The SAR concluded tunnel nitrogen spills were under-investigated and led to this work

  5. 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.

  6. Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 538: Spill Sites, Nevada Test Site, Nevada, Rev. No.: 0

    Energy Technology Data Exchange (ETDEWEB)

    Alfred Wickline

    2006-04-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the actions necessary for the closure of Corrective Action Unit (CAU) 538: Spill Sites, Nevada Test Site, Nevada. It has been developed in accordance with the ''Federal Facility Agreement and Consent Order'' (FFACO) (1996) that was agreed to by the State of Nevada, the U.S. Department of Energy (DOE), and the U.S. Department of Defense. A SAFER may be performed when the following criteria are met: (1) Conceptual corrective actions are clearly identified (although some degree of investigation may be necessary to select a specific corrective action before completion of the Corrective Action Investigation [CAI]). (2) Uncertainty of the nature, extent, and corrective action must be limited to an acceptable level of risk. (3) The SAFER Plan includes decision points and criteria for making data quality objective (DQO) decisions. The purpose of the investigation will be to document and verify the adequacy of existing information; to affirm the decision for either clean closure, closure in place, or no further action; and to provide sufficient data to implement the corrective action. The actual corrective action selected will be based on characterization activities implemented under this SAFER Plan. This SAFER Plan identifies decision points developed in cooperation with the Nevada Division of Environmental Protection (NDEP) and where DOE will reach consensus with NDEP before beginning the next phase of work.

  7. 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.

  8. Vegetable oil spills : oil properties and behaviour

    International Nuclear Information System (INIS)

    Fingas, M.; Fieldhouse, B.; Jokuty, P.

    2001-01-01

    In 1997, the United States Environmental Protection Agency conducted a thorough review of the issue regarding vegetable oil spills. Recent attention has refocused on this issue as a result of an incident where 20 tons of canola oil was spilled in the Vancouver Harbour in 2000. In the past, vegetable oils were suggested to be a useful test material because they were thought to be innocuous. It was even suggested they be used to remove petroleum oil residues from beaches. However, recent studies have shown that spills of vegetable oils can have major environmental consequences, equivalent to those of petroleum oil spills. The spills have devastating effects on birds and intertidal organisms. This paper presented a summary of historical vegetable spills from around the world. In this study, specific behaviour tests were examined for several oils including canola, soy bean, olive, castor and corn oils. Evaporation, water-in-oil emulsification and chemical dispersion were measured and were found to be nearly zero, suggesting that vegetable oil spills are not very soluble in water. The aquatic toxicity of vegetable oil is low, but their fate is quite different from petroleum. Vegetable oils do not evaporate to a significant degree, they do not form water-in-oil emulsions, nor do they disperse in water. The physical properties of vegetable oils were also measured, including density and viscosity. This paper presented the aquatic toxicity of several vegetable oils along with other environmental data including the degradation rates noted in the literature. Most environmental damage reported in the literature is by contact with birds feathers resulting in hypothermia and secondly by smothering of intertidal organisms. The effect of vegetable oil on fish has not been well studied, but it is expected that there will be little destructive effect except where smothering can occur. 35 refs., 3 tabs

  9. Vegetable oil spills : oil properties and behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Fingas, M.; Fieldhouse, B.; Jokuty, P. [Environment Canada, Ottawa, ON (Canada). Emergencies Science Div

    2001-07-01

    In 1997, the United States Environmental Protection Agency conducted a thorough review of the issue regarding vegetable oil spills. Recent attention has refocused on this issue as a result of an incident where 20 tons of canola oil was spilled in the Vancouver Harbour in 2000. In the past, vegetable oils were suggested to be a useful test material because they were thought to be innocuous. It was even suggested they be used to remove petroleum oil residues from beaches. However, recent studies have shown that spills of vegetable oils can have major environmental consequences, equivalent to those of petroleum oil spills. The spills have devastating effects on birds and intertidal organisms. This paper presented a summary of historical vegetable spills from around the world. In this study, specific behaviour tests were examined for several oils including canola, soy bean, olive, castor and corn oils. Evaporation, water-in-oil emulsification and chemical dispersion were measured and were found to be nearly zero, suggesting that vegetable oil spills are not very soluble in water. The aquatic toxicity of vegetable oil is low, but their fate is quite different from petroleum. Vegetable oils do not evaporate to a significant degree, they do not form water-in-oil emulsions, nor do they disperse in water. The physical properties of vegetable oils were also measured, including density and viscosity. This paper presented the aquatic toxicity of several vegetable oils along with other environmental data including the degradation rates noted in the literature. Most environmental damage reported in the literature is by contact with birds feathers resulting in hypothermia and secondly by smothering of intertidal organisms. The effect of vegetable oil on fish has not been well studied, but it is expected that there will be little destructive effect except where smothering can occur. 35 refs., 3 tabs.

  10. 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.

  11. 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.

  12. 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)

  13. 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.

  14. 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

  15. 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.

  16. 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.)

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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)

  2. When oil spills emulsify

    International Nuclear Information System (INIS)

    Bobra, M.; Fingas, M.; Tennyson, E.

    1992-01-01

    Cleanup operations of oil spills must take into account the numerous detrimental effects attributable to the emulsification of spilled oil into a stable water-in-oil mousse. The incorporation of water greatly increases the volume of the polluted material. The viscous nature of mousse impedes the efficient operation of most mechanical recovery equipment and results in a cohesive slick that resists dispersion, both natural and artificial. The rate at which spilled oil emulsifies determines the effective window of opportunity for specific countermeasures. Much has been learned from previous studies on petroleum emulsification, but is still remain a poorly understood phenomenon. Although most crude oils can be emulsified, not all spills result in the formation of stable mousse. The formation of mousse results from a complex series of processes. Whether an oil will form mousse or not, and if so, at what rate, depends on an array of different factors including the properties of the oil and the prevailing environmental conditions. We need a greater understanding of the emulsification process to better predict the emulsification behavior of oil spills and utilize the most appropriate countermeasures available. In this paper, the authors report on work to elucidate the role that physicochemical factors play in determining an oil's tendency to emulsify. The authors studied the emulsification behavior of oils of known composition to examine the importance of oil chemistry in the emulsification process

  3. 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...

  4. 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...

  5. Proceedings of the nineteenth arctic and marine oil spill program (AMOP) technical seminar

    International Nuclear Information System (INIS)

    1996-01-01

    The technical seminar on arctic and marine oil spills introduced issues concerning oil spills at sea, in particular the critical first few hours of an oil spill. State-of-the-art technologies which assist the response team in tracking and predicting the behavior of oil spills, were described. The physical and chemical properties of spilled oil were studied, as well as those of oil spill treating agents, including testing their biological effects. New methods to contain and recover spilled oil were reviewed. Volume 2 presented results from field experiments in which in-situ burning was performed, and demonstrated modelling techniques for the detection, prediction and tracking of oil spills. refs., tabs., figs

  6. Concrete protection from sodium spills by intentionally defected liners, small-scale tests S9 and S10

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Boehmer, W.D.

    1975-07-01

    Two small scale tests were performed to determine the protection against sodium attack afforded to a concrete surface by a defected steel liner. An inert atmosphere was maintained over the sodium pool, which was heated electrically to 1600 0 F for 2--6 hrs in one test, to 1380 0 F for 19 hrs in the other. The 10 inch diameter vertical concrete surface was separated from the sodium by a liner plate in which small defects had been drilled. The plates provided significant protection against direct chemical attack, but most of the water was released from the concrete through the defects to react in the sodium pool region. The liners were corroded significantly in the defect areas. (U.S.)

  7. 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

  8. 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

  9. Spill reporting and prevention

    International Nuclear Information System (INIS)

    Swiss, J.J.

    1997-01-01

    The actions that companies in British Columbia are required to take to comply with spill reporting requirements and with the waste management legislation of the B.C. Waste Management Act were discussed. A company's ability to respond effectively to hazardous materials spills depends on three factors: (1) understanding the regulatory requirements, (2) having an emergency response capability, and (3) having a staff trained to exercise those responsibilities. The steps involved in complying with the legislation were outlined . The types and quantities of spilled material that must be reported were listed, and advice was given on how a company can effectively incorporate emergency planning into its Environmental Health and Safety Management System. Responsibilities of the the individual designated as the on-scene commander were also spelled out. 3 tabs

  10. Oil spill response plan

    International Nuclear Information System (INIS)

    1999-08-01

    The plan outlined in this document specifies the actions that the Canadian Wildlife Service Atlantic Region is mandated to take in the event of an oil spill, or on discovering oiled migratory birds in terrestrial, fresh water, marine and inter-tidal habitats. In addition to describing the role and responsibilities of the Canadian Wildlife Service, the document also describes response plans of other agencies for dealing with all wildlife species affected by oil spills. Reporting paths, the lead agency concept, shared responsibilities with other Canadian Wildlife Service regional offices, provincial agencies, Heritage Canada, non-government wildlife response agencies, oil spill response organizations, and international organizations are outlined. An overview of the reporting and communications process is also provided

  11. 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.

  12. 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

  13. 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)

  14. 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.

  15. Spill response trade-offs in a very large spill

    International Nuclear Information System (INIS)

    Schulze, R.

    1990-01-01

    This paper examines the physical limitaions on spill encounter rate and how these limitaions affect the are that can be covered in spill response. Since mechanical recovery devices may not be able to cover the affected area in a large spill, in situ burning must be considered as a response option. Further, effective recovered oil logistics is essential to successful response operations and keeping skimmers operating. A successful spill response effort in a very large oil spill often depends on: Prompt response, Skimmer encounter rate, Making a decision for in situ burning, Recovered oil logistics

  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. 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

  18. Spill response : an exercise in teamwork

    International Nuclear Information System (INIS)

    Anon.

    1998-01-01

    An offshore oil spill response exercise was conducted at Hibernia to demonstrate to the Canada-Newfoundland Offshore Petroleum Board the emergency response capabilities that are in place in the event of large offshore spills. The Canadian Coast Guard, Eastern Canada Response Corporation Ltd., Hibernia, Husky Oil Operations Ltd., Jeanne d'Arc Basin Operators Group and the Terra Nova Project team participated in the exercise. The exercise was a success in that it demonstrated that the emergency response teams have the capability of containing and recovering large and small offshore oil spills. The two systems that were tested during the exercise were the large wide-swath boom system and a smaller side-sweep system. Two supply vessels worked in tandem. 11 figs

  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. U.S. Coast Guard oil spill remote sensing : preliminary laser fluorosensor studies

    International Nuclear Information System (INIS)

    Fant, J.W.; Hansen, K.A.

    2005-01-01

    Maritime oil spill events are costly and damaging to the environment. Nearly 40 per cent of ship sourced spills occurring in the last 25 years have involved medium to heavy grade fuel oils. There is, therefore, an immediate need to detect and track subsurface oil spills, particularly as heavy and weathered oil can sink below the surface during a spill and often becomes problematic to detect, track and recover. The United States Coast Guard has limited capabilities to detect and track an oil spill, especially in poor weather. This paper discussed research and assessment efforts focused on laser fluorosensor technology. Testing of 3 independent laser fluorosensing systems was conducted to determine sensing depth capabilities and sensor shortcomings in ideal conditions. Studies included the detection and collection of laser induced fluorescence spectra at the surface as well as at various depths down to 5 metres in both daylight and night-time environments. The sensors were tested to assess their capabilities to meet the Coast Guard's oil sensor and operational requirements. Three sensors were tested by the Coast Guard at the Ohmsett National Oil Response Test Facility: the Airborne Oceanographic Lidar (AOL-3), a light detection and ranging system (lidar) to measure biological and physical oceanographic features developed by the National Aeronautics and Space Administration (NASA); the Fluorescent Lidar Spectrometer (FLS) lidar, developed by Laser Diagnostic Instruments International Inc. of Canada; and the Ultraviolet Biological Trigger Lidar, developed by Science and Engineering Services, Inc. (SESI) to detect and discriminate bio-warfare agent aerosols for the United States Army. The 3 fluorometers exhibited the ability to detect oil both on and below the water's surface. There were differences in the peak locations in the spectrum for the same oils among the lasers tested. It was also noted that all the systems had the capability of detecting oil in a night

  3. 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)

  4. Assessment and recommendations for the oil spill cleanup of Guanabara Bay, Brazil

    International Nuclear Information System (INIS)

    Michel, Jacqueline

    2000-01-01

    The potential impact of an oil spill consisting of a blend of diesel and a heavy residual fuel oil on water column resources, benthic resources, intertidal habitats and communities, birds, and socio-economic resources was assessed and recommendations for clean-up given. The oil spill occurred in the bay on January 18, 2000

  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. Aerosols generated by spills of viscous solutions and slurries

    International Nuclear Information System (INIS)

    Ballinger, M.Y.; Hodgson, W.H.

    1986-12-01

    Safety assessments and environmental impact statements for nuclear fuel cycle facilities require an estimate of potential airborne releases caused by accidents. Aerosols generated by accidents are being investigated by Pacific Northwest Laboratory to develop methods for estimating source terms from these accidents. Experiments were run by spilling viscous solutions and slurries to determine the mass and particle-size distribution of the material made airborne. In all cases, 1 L of solution was spilled from a height of 3 m. Aqueous solutions of sucrose (0 to 56%) gave a range of viscosities from 1.3 to 46 cp. The percent of spill mass made airborne from the spills of these solutions ranged from 0.001 to 0.0001. The mass of particles made airborne decreased as solution viscosity increased. Slurry loading ranged from 25 to 51% total solids. The maximum source airborne (0.0046 wt %) occurred with the slurry that had the lightest loading of soluble solids. The viscosity of the carrying solution also had an impact on the source term from spilling slurries. The effect of surface tension on the source term was examined in two experiments. Surface tension was halved in these spills by adding a surfactant. The maximum weight percent airborne from these spills was 0.0045, compared to 0.003 for spills with twice the surface tension. The aerodynamic mass medium diameters for the aerosols produced by spills of the viscous solutions, slurries, and low surface tension liquids ranged from 0.6 to 8.4 μm, and the geometric standard deviation ranged from 3.8 to 28.0

  8. Oil spill recovery technology

    International Nuclear Information System (INIS)

    Nash, J.; Cooper, W.; Nee, V.; Nigim, H.

    1992-01-01

    Current deficiencies in oil spill cleanup processes have resulted in research and development of new cleanup technologies at the University of Notre Dame. Emphasis on reducing, reusing and recycling equipment and waste at a cleanup site has prompted advances in oil recovery technology as well as improvement in sorbent materials. (author)

  9. 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

  10. 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.

  11. 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

  12. 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)

  13. 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)

  14. 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...

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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).

  20. 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)

  1. Japan FRI research activities on oil tank/spilled oil fire

    International Nuclear Information System (INIS)

    Koseki, Hiroshi

    1992-01-01

    Introduction of research activities on oil tank/spilled oil fire at FRI, Japan is done. FRI has a long history of studying oil tank and spilled oil fires. Many large oil fire tests were done. The purpose of these studies is different with research of response of oil spill, but the accumulation of this knowledge is useful for conducting elimination of spilled oil on the sea with burning. Therefore to do collaboration with fire science research groups, such as FRI is useful for future activities for response to oil spills

  2. 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

  3. 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

  4. 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.

  5. 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)

  6. 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)

  7. 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.

  8. Guide to oil spill exercise planning

    International Nuclear Information System (INIS)

    1996-01-01

    The International Convention on Oil Pollution Preparedness, Response and Cooperation (OPRC Convention) foresees a future in which all at risk states have national oil spill preparedness and response plans. The Convention also encourages the idea that national plans be developed in cooperation with oil and shipping industries. The ultimate test of any contingency plan is measured by performance in a real emergency. It is vital, therefore, that any programme for developing a national contingency plan must include an ongoing programme to test the plan through realistic exercises. An exercise programme must progressively prepare the Oil Spill Energy Response Team to perform effectively in realistic representations of the risks that the contingency plan has been designed to meet. This report has been designed to guide all those in government or industry who are faced with the responsibility of developing and managing oil spill response exercises at all levels. It carries with it the authority that derives from peer review by many centres of oil spill response excellence around the world. It is well-illustrated with brief case histories of exercises that have been carried out by many IPIECA member companies. Each of those companies has indicated its preparedness to share more information by providing contact name and address details within this report. (author)

  9. 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)

  10. 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.

  11. 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)

  12. 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)

  13. 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

  14. 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

  15. Oil well spill trough

    International Nuclear Information System (INIS)

    Wigington, J.R. Sr.

    1992-01-01

    This patent describes a process involving an oil well and rig having a casing, a platform on the rig extending around the casing. This patent describes improvement in pulling the tubing from the casing; disconnecting joints of tubing thereby; and spilling liquids from the casing, catching spilled liquids from the casing in a basin below the platform, draining the basin substantially simultaneously; connecting the drain hole to a tank, and reducing the pressure in the tank to less than atmospheric pressure. This paper also describes an oil well and rig having a casing; the rig having a platform extending around the casing. This patent describes improvement in a basin surrounding the casing and connected thereto, the basin below the platform, a drain connection in the lower part of the basin, a conduit connected to the drain, and means for applying a suction to the conduit

  16. 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)

  17. 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

  18. 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.

  19. 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.

  20. At spille en krimi

    DEFF Research Database (Denmark)

    Sandvik, Kjetil

    2008-01-01

    væsentlige forskel at vi her ikke længere er læsere eller seere, men aktive medspillere i selve opklaringsarbejdet som ledere af eller deltagere i et efterforskningshold. Denne artikel ser nærmere på den interaktive krimi som performativ fortælling, dvs. en fortælling som bliver til i kraft af at den spilles...

  1. 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

  2. Dam spills and fishes

    International Nuclear Information System (INIS)

    1996-01-01

    This short paper reports the main topics discussed during the two days of the annual colloquium of the Hydro-ecology Committee of EdF. The first day was devoted to the presentation of the joint works carried out by EdF, the Paul-Sabatier University (Toulouse), the Provence St-Charles University (Marseille), the ENSAT (Toulouse) and the CEMAGREF (Lyon and Aix-en-Provence) about the environmental impact of dam spills on the aquatic flora and fauna downstream. A synthesis and recommendations were presented for the selection and characterization of future sites. The second day was devoted to the hydro-ecology study of the dam reservoir of Petit-Saut (French Guyana): water reoxygenation, quality evolution, organic matter, plankton, invertebrates and fishes. The 134 French dams concerned by water spills have been classified according to the frequency of spills, the variations of flow rates created, and their impacts on fishing, walking, irrigation, industry, drinking water, navigation, bathing. Particular studies on different sites have demonstrated the complexity of the phenomena involved concerning the impact on the ecosystems and the water quality. (J.S.)

  3. 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)

  4. 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

  5. Characterization and identification of Detroit River mystery oil spill (2002)

    International Nuclear Information System (INIS)

    Wang, Z.; Fingas, M.; Lambert, P.

    2003-01-01

    The authors described the mysterious oil spill which occurred in the Detroit River in 2002. Advanced chemical fingerprinting and data interpretation techniques were conducted on spill samples collected by Environment Canada, Ontario Region, to determine the chemical composition of the oil and find out where it came from. The objective was to gather information concerning the nature, type, and components of the spill samples. The authors checked if the samples were identical to determine if they originated from the same source. They used a tiered analytical approach which facilitates the detailed compositional analysis by gas chromatograph-mass spectrometer (GC-MS) and GC-flame ionization detection (FID). A wide range of diagnostic ratios of source-specific marker compounds for interpreting chemical data was determined and analyzed. The results proved that: (1) the spill samples were largely composed of lube oil mixed with a smaller portion of diesel fuel, (2) sample number 3 collected from N. Boblo Island was more weathered than samples 1 and 2, (3) the oil in three samples was the same and originated from the same source, as shown by fingerprinting results, (4) most PAH compounds were from the diesel portion in the spill samples, and the biomarker compounds were mostly from the lube oil, (5) the diesel in the samples had been weathered and degraded, and the lube oil in the spill samples was waste lube oil, and (6) input of pyrogenic PAHs to the spill samples was clearly proven. The spill likely came from a place where both combustion and motor lubrication processes occur. 46 refs., 4 tabs., 6 figs

  6. Quality assurance plan for the data acquisition and management system for monitoring the fuel oil spill at the Sandia National Laboratories installation in Livermore, California

    International Nuclear Information System (INIS)

    Peerenboom, J.P.; Leser, C.C.; Ramsey, G.M.; Widing, M.A.

    1995-04-01

    In February 1975, the accidental puncture of an underground transfer line buried about 4 ft below the ground surface at the SNL installation in Livermore, California, resulted in the release of approximately 225.5 m 3 of No. 2 diesel fuel. This report describes the formal quality assurance plan that will be used for the data acquisition and management system developed to monitor a bioremediation pilot study by Argonne National Laboratory in association with Sandia National Laboratories. The data acquisition and management system will record the site data during the bioremediation effort and assist users in site analysis. The designs of the three major subsystems of this system are described in this report. Quality assurance criteria are defined for the management, performance, and assessment of the system. Finally, the roles and responsibilities for configuration management of this system are defined for the entire life cycle of the project

  7. 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.'

  8. 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...

  9. 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

  10. 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).

  11. 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)

  12. On-site fuel cell field test support program

    Science.gov (United States)

    Staniunas, J. W.; Merten, G. P.

    1982-01-01

    In order to assess the impact of grid connection on the potential market for fuel cell service, applications studies were conducted to identify the fuel cell operating modes and corresponding fuel cell sizing criteria which offer the most potential for initial commercial service. The market for grid-connected fuel cell service was quantified using United's market analysis program and computerized building data base. Electric and gas consumption data for 268 buildings was added to our surveyed building data file, bringing the total to 407 buildings. These buildings were analyzed for grid-isolated and grid-connected fuel cell service. The results of the analyses indicated that the nursing home, restaurant and health club building sectors offer significant potential for fuel cell service.

  13. A nondestructive testing device for determining 235U enrichment in power reactor fuel elements

    International Nuclear Information System (INIS)

    Liu Lanhua; Liu Nangai

    1990-07-01

    The development and application of a nondestructive testing device are presented, which is used for determining the 235 U enrichment in the mixed fuel of fuel elements with UO 2 pellets. The testing efficiency is improved because the passive gamma ray method and a hole-bored NaI crystal and four channel multichannel analyzer are used. The false discrimination rate is reduced as the average comparing method is taken. This device is simple in structure and easy in operation. It has provided a new testing tool for the fuel elements production in China. This device has successfully been used in Qinshan Nuclear Power Plant in testing its fuel elements

  14. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    International Nuclear Information System (INIS)

    Grover, S.B.

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  15. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  16. Performance evaluation of UO2-Zr fuel in power ramp tests

    International Nuclear Information System (INIS)

    Knudsen, P.; Bagger, C.

    1977-01-01

    In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

  17. The development of the neutron flux measurement technology using SPNDs during nuclear fuel irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B. G.; Kang, Y. H.; Cho, M. S.; Joo, K. N.; Choi, M. H.; Park, S. J.; Shin, Y. T.; Oh, J. M.; Kim, Y. J

    2004-03-01

    As a part of the development of instrumentation technologies for a nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), a study is performed to measure and evaluate the neutron flux at the same position as the nuclear fuel during irradiation test using the SPND(Self Powered Neutron Detector). To perform this study, rhodium type SPNDs and amplifier are selected suitable to irradiation test, and the selected SPNDs are installed in instrumented fuel capsule(02F-11K). The irradiation test using a instrumented fuel capsule are performed in the OR5 vertical hole of HANARO for about 54 days, and SPND output signals are acquired successfully during irradiation test. Acquired SPND signals are analyzed and evaluated as a reliable data by COSMOS Code. This will be utilized for the fuel related research together with fuel center temperature and reactor operation data.

  18. Performance Degradation Tests of Phosphoric Acid Doped PBI Membrane Based High Temperature PEM Fuel Cells

    DEFF Research Database (Denmark)

    Zhou, Fan; Araya, Samuel Simon; Grigoras, Ionela

    2014-01-01

    Degradation tests of two phosphoric acid (PA) doped PBI membrane based HT-PEM fuel cells were reported in this paper to investigate the effects of start/stop and the presence of methanol in the fuel to the performance degradation. Continuous tests with H2 and simulated reformate which was composed...... of H2, water steam and methanol as the fuel were performed on both single cells. 12-h-startup/12-h-shutdown dynamic tests were performed on the first single cell with pure dry H2 as the fuel and on the second single cell with simulated reformate as the fuel. Along with the tests electrochemical...... techniques such as polarization curves and electrochemical impedance spectroscopy (EIS) were employed to study the degradation mechanisms of the fuel cells. Both single cells showed an increase in the performance in the H2 continuous tests, because of a decrease in the ORR kinetic resistance probably due...

  19. Pigouvian penalty for oil spills

    International Nuclear Information System (INIS)

    Kohn, R.E.

    1993-01-01

    The imposition of ex ante taxes on expected spilled oil, in addition to ex post payments for damages under tort liability, would foster economic efficiency. This paper begins the analysis of the joint approach with the case in which Pigouvian taxes are used exclusively. A model is developed in which the volume of spilled oil causing environmental damage is reduced, first by spill prevention expenditures by shippers and then by clean-up expenditures by the government. The efficient Pigouvian tax on expected spilled oil equals marginal environmental damage which equals the net marginal cost of prevention which equals marginal clean-up cost. (Author)

  20. CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 204: STORAGE BUNKERS, NEVADA TEST SITE, NEVADA

    International Nuclear Information System (INIS)

    2006-01-01

    Corrective Action Unit (CAU) 330 consists of four Corrective Action Sites (CASs) located in Areas 6, 22, and 23 of the Nevada Test Site (NTS). The unit is listed in the Federal Facility Agreement and Consent Order (FFACO, 1996) as CAU 330: Areas 6, 22, and 23 Tanks and Spill Sites. CAU 330 consists of the following CASs: CAS 06-02-04, Underground Storage Tank (UST) and Piping CAS 22-99-06, Fuel Spill CAS 23-01-02, Large Aboveground Storage Tank (AST) Farm CAS 23-25-05, Asphalt Oil Spill/Tar Release

  1. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Min, Sohn Jae; Kang, Y. H.; Kim, B. G. [and others

    2001-11-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO, the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT. The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. The out-of-pile test system for pressure measurement was developed, and the test with the LVDT at room temperature(19 .deg. C) were performed. A out-of-pile test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2} and repeated 6 times at each condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. This report describes the system configuration, the out-of-pile test procedures, and the results. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics for the detail design of the fuel irradiation capsule.

  2. Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

    International Nuclear Information System (INIS)

    Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

    1993-01-01

    The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO 2 pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel

  3. Post-test thermomechanical calulations and preliminary data analysis for the Spent Fuel Test: Climax

    International Nuclear Information System (INIS)

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

    1985-09-01

    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. Thermomechanical response of the SFT-C was calculated before the test began using the finite-element structural analysis code ADINA and its companion heat transfer code ADINAT. While we found that the level of agreement between measured and calculated rock displacements was quite good, we needed to revise certain aspects of the heat transfer calculation, material properties, and in situ stresses to incorporate information obtained during and after the heated phase of the test. The post-test calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels that were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares those results with selected measurements made during the 3-year heating phase and 6-month cooling phase of the SFT-C

  4. Analysis of the SBLOCAs in HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-09-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss Of Coolant Accidents (SBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperatures (PCT) are predicted to be about 906.9 .deg. C for the cold leg break accident in PWR fuel test mode and 971.9 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 6% of the cross section area of the pipe for PWR fuel test mode and the 8% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  5. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  6. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Joo, K. N.; Park, S. J.; Kang, Y. H.; Kim, Y. K.; Yeum, K. I. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. Therefore, the out of pile test system for pressure measurement was developed, and the test with the LVDT at room temperature were performed. This test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2}, and repeated 6 times at same condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule.

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

    Science.gov (United States)

    1984-10-01

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

  8. Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doherty, A.L.

    1997-06-01

    This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs

  9. Engineering study: Fast Flux Test Facility fuel reprocessing

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  10. Back pressure helium leak testing of fuel elements for Dhruva research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, N G; Ahmad, Anis; Kulkarni, P G; Purushotham, D S.C. [Bhabha Atomic Research Centre, Bombay (India). Atomic Fuels Div.

    1994-12-31

    Leak tightness specification on fuel elements for reactor use is always very stringent. The fuel element fabricated for Dhruva reactor is specified to be leak-tight up to 1 x 10{sup -8} std. cc/sec. The fuel element consists of natural metallic uranium rod around 12.5 mm diameter and 3 meter long in encased in aluminium tube and seal welded at both ends. Since helium gas is not filled inside the fuel element while doing seal welding, the only way to do helium leak testing of such fuel rods is by back-pressure technique. This paper describes the development of test facility for carrying out such test and discusses the experiences of carrying out helium leak testing by back-pressure technique on more than 700 numbers of fuel rods for Dhruva reactor. (author). 4 refs., 3 figs., 1 tab.

  11. Analysis of th SBLOCAs in the room 1 for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-10-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss of Coolant Accidents (SBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperature (PCT) is predicted to be about 931.4 .deg. C for the cold leg break accident in PWR fuel test mode and 931.6 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 8% of the cross section area of the pipe for PWR fuel test mode and the 10% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  12. First TREAT transient overpower tests on U-Pu-Zr fuel: M5 and M6

    International Nuclear Information System (INIS)

    Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

    1987-01-01

    Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) test of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the integral fast reactor concept. Similar tests on U-5 wt% Fs fueled Experimental Breeder Reactor (EBR)-II driver pins were previously performed and reported. Results from these earlier tests indicated a margin to failure of ∼ 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuels types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety-related fuel performance characteristics of U-Fs and U-Pu-Zr. Tests M5 and M6 indicate that, under the TOP conditions used in the tests, ternary fuel displayed about the same margin to failure as U-Fs fuel. At low burnups, ternary fuel showed less prefailure axial elongation than observed in U-Fs pins, but elongations of 3 to 5% might turn out to be typical. Finally, fuel from the breached ternary pin in M6 showed, qualitatively, the same benignly dispersive behavior as U-Fs

  13. Respiratory Effects of the Hebei Spirit Oil Spill on Children in Taean, Korea

    OpenAIRE

    Jung, Suk-Chul; Kim, Kyung-Mook; Lee, Kun-Song; Roh, Sangchul; Jeong, Woo-Chul; Kwak, Sahng-June; Lee, Ik-Jin; Choi, Young-Hyun; Noh, Su Ryeon; Hur, Jong-Il; Jee, Young-Koo

    2013-01-01

    Purpose The oil spill from the Heibei Spirit in December 2007 contaminated the Yellow Coast of South Korea. We evaluated the respiratory effects of that spill on children who lived along the Yellow Coast. Methods Of 662 children living in the area exposed to the oil spill, 436 (65.9%) were enrolled as subjects. All subjects completed a modified International Study of Asthma and Allergies in Childhood questionnaire. A health examination, including a skin prick test, pulmonary function test, an...

  14. Performance test of the I and C system (GSF - 2002) for the irradiation tests using a fuel capsule

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Park, S. J.; Kim, B. G.; Ahn, D. H

    2004-12-01

    HANARO is a very important facility in Korea. It offers various types of irradiation tests of nuclear fuels and materials. With the various applications of the HANARO capsule for the academic and industrial applications, new technologies and relevant facilities will become more important especially for the advanced nuclear fuels and materials development. A new I and C system for an irradiation test using an instrumented fuel capsule have been designed and manufactured to provide more qualified data to fuel developer. The performance test which started in 2004, was done to investigate the thermal response of the capsule connected to the gas mixing system of the new I and C system(GSF-2002) in the cold test loop under the HANARO hydraulic operational condition. Main test parameters are mass flow rate of 25, 50 and 100 cc/min of He/Ne gas, gas pressure of 1 to 3 kg/cm{sup 2}, heater power of 1 to 3.4kW and different gas mixing ratios of He to Ne. From the out-pile tests, it was confirmed that the I and C system(GSF-2002) would be feasible for the fuel irradiation tests. Both analytical and test data prepared by this study are directly used for the fuel experiments related to advanced fuel development program.

  15. On-Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Hawari, Ayman I.; Bourham, Mohamed A.

    2010-01-01

    Very High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (∼ 1-mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4%-10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  16. Survey of European LWR fuel irradiation test facilities

    Energy Technology Data Exchange (ETDEWEB)

    Hardt, P von der [Commission of the European Communities, Joint Research Centre, Petten Establishment, Petten (Netherlands)

    1983-06-01

    The first European commercial nuclear power plants (1956) featured gas-cooled thermal reactors. Although there is now a general orientation towards light water cooled plants (with a slight preference for the PWR) a large fraction of the 1982 nuclear generating capacity is still invested in gas-cooled reactors. R and D also continues for the HTGR with its long-term development potential. This paper, however, is limited to a general survey of experimental programmes and facilities for light water reactor fuel testing in Western Europe, particularly inside the European Communities. As it turns out, over a dozen major installations are available, all connected to research reactors in government-funded R and D centres. Their equipment is briefly reviewed. Some 50% of the experimental programmes are carried out in large international collaboration, involving up to 20 organizations per project. Techniques and results are rapidly communicated through frequent meetings and conferences. It is anticipated that a part of the present research reactor-based work will gradually shift to power reactor pool side inspection facilities. (author)

  17. Survey of European LWR fuel irradiation test facilities

    International Nuclear Information System (INIS)

    Hardt, P. von der

    1983-01-01

    The first European commercial nuclear power plants (1956) featured gas-cooled thermal reactors. Although there is now a general orientation towards light water cooled plants (with a slight preference for the PWR) a large fraction of the 1982 nuclear generating capacity is still invested in gas-cooled reactors. R and D also continues for the HTGR with its long-term development potential. This paper, however, is limited to a general survey of experimental programmes and facilities for light water reactor fuel testing in Western Europe, particularly inside the European Communities. As it turns out, over a dozen major installations are available, all connected to research reactors in government-funded R and D centres. Their equipment is briefly reviewed. Some 50% of the experimental programmes are carried out in large international collaboration, involving up to 20 organizations per project. Techniques and results are rapidly communicated through frequent meetings and conferences. It is anticipated that a part of the present research reactor-based work will gradually shift to power reactor pool side inspection facilities. (author)

  18. The INCOTUR model : estimation of losses in the tourism sector in Alcudia due to a hydrocarbon spill

    International Nuclear Information System (INIS)

    Bergueiro, J.R.; Moreno, S.; Guijarro, S.; Santos, A.; Serr, F.

    2006-01-01

    This paper presented a computer model that calculates the economic losses incurred by a hydrocarbon spill on a coastal area. In particular, it focused on the Balearic Islands in the Bay of Alcudia where the economy depends mainly on tourism. A large number of oil tankers carrying crude oil and petroleum products pass through the Balearic Sea. Any pollution resulting from a fuel spill can have a significant economic impact on both the tourism sector and the Balearic society in general. This study focused on the simulation of 18 spills of Jet A1 fuel oil, unleaded gasoline and Bunker C fuel oil. Simulations of the study area were produced with OILMAP, MIKE21, GNOME and ADIOS models which estimated the trajectories of various spills and the amount of oil washed ashore. The change in physical and chemical properties of the spilled hydrocarbons was also determined. The simulation models considered the trajectory followed by spills according to the type and amount of spill, weather conditions prevailing during the spill and the period immediately following the spill. The INCOTUR model was then used to calculate the economic losses resulting from an oil spill by considering the number of tonnes of oil washed ashore; number of days needed to organize cleanup; the percentage of tourism that will be maintained despite the effects of the spill; number of hotel beds; percentage of hotel occupancy by month; cost of package holidays; petty cash expenses; and, cost of advertising campaign for the affected area. With this data, the model can determine the number of days needed to clean and restore the coastline; monthly rate of recovery in tourism levels; and, losses in tourism sector. According to the INCOTUR model, the total losses incurred by a spill of 40,000 tonnes of Bunker C fuel, was estimated at 472 million Euros. 9 refs., 2 tabs., 12 figs

  19. The INCOTUR model : estimation of losses in the tourism sector in Alcudia due to a hydrocarbon spill

    Energy Technology Data Exchange (ETDEWEB)

    Bergueiro, J.R.; Moreno, S.; Guijarro, S.; Santos, A.; Serr, F. [Iles Balears Univ., Palma de Mallorca, Balearic Islands (Spain). Dept. of Chemistry

    2006-07-01

    This paper presented a computer model that calculates the economic losses incurred by a hydrocarbon spill on a coastal area. In particular, it focused on the Balearic Islands in the Bay of Alcudia where the economy depends mainly on tourism. A large number of oil tankers carrying crude oil and petroleum products pass through the Balearic Sea. Any pollution resulting from a fuel spill can have a significant economic impact on both the tourism sector and the Balearic society in general. This study focused on the simulation of 18 spills of Jet A1 fuel oil, unleaded gasoline and Bunker C fuel oil. Simulations of the study area were produced with OILMAP, MIKE21, GNOME and ADIOS models which estimated the trajectories of various spills and the amount of oil washed ashore. The change in physical and chemical properties of the spilled hydrocarbons was also determined. The simulation models considered the trajectory followed by spills according to the type and amount of spill, weather conditions prevailing during the spill and the period immediately following the spill. The INCOTUR model was then used to calculate the economic losses resulting from an oil spill by considering the number of tonnes of oil washed ashore; number of days needed to organize cleanup; the percentage of tourism that will be maintained despite the effects of the spill; number of hotel beds; percentage of hotel occupancy by month; cost of package holidays; petty cash expenses; and, cost of advertising campaign for the affected area. With this data, the model can determine the number of days needed to clean and restore the coastline; monthly rate of recovery in tourism levels; and, losses in tourism sector. According to the INCOTUR model, the total losses incurred by a spill of 40,000 tonnes of Bunker C fuel, was estimated at 472 million Euros. 9 refs., 2 tabs., 12 figs.

  20. Post-test investigation result on the WWER-1000 fuel tested under severe accident conditions

    International Nuclear Information System (INIS)

    Goryachev, A.; Shtuckert, Yu.; Zwir, E.; Stupina, L.

    1996-01-01

    The model bundle of WWER-type were tested under SFD condition in the out-of-pile CORA installation. The objective of the test was to provide an information on the WWER-type fuel bundles behaviour under severe fuel damage accident conditions. Also it was assumed to compare the WWER-type bundle damage mechanisms with these experienced in the PWR-type bundle tests with aim to confirm a possibility to use the various code systems, worked our for PWR as applied to WWER. In order to ensure the possibility of the comparison of the calculated core degradation parameters with the real state of the tested bundle, some parameters have been measured on the bundle cross-sections under examination. Quantitative parameters of the bundle degradation have been evaluated by digital image processing of the bundle cross-sections. The obtained results are shown together with corresponding results obtained by the other participants of this investigation. (author). 3 refs, 13 figs

  1. Verification tests for CANDU advanced fuel -Development of the advanced CANDU technology-

    International Nuclear Information System (INIS)

    Chung, Jang Hwan; Suk, Ho Cheon; Jeong, Moon Ki; Park, Joo Hwan; Jeong, Heung Joon; Jeon, Ji Soo; Kim, Bok Deuk

    1994-07-01

    This project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year Out-of-pile hydraulic tests for the prototype of CANFLEX bundle was conducted in the CANDU-hot test loop at KAERI. Thermalhydraulic analysis with the assumption of CANFLEX-NU fuel loaded in Wolsong-1 was performed by using thermalhydraulic code, and the thermal margin and T/H compatibility of CANFLEX bundle with existing fuel for CANDU-6 reactor have been evaluated. (Author)

  2. Fabrication of Non-instrumented capsule for DUPIC simulated fuel irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Kang, Y.H.; Park, S.J.; Shin, Y.T. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    In order to develope DUPIC nuclear fuel, the irradiation test for simulated DUPIC fuel was planed using a non-instrumented capsule in HANARO. Because DUPIC fuel is highly radioactive material the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO was designed to remotely assemble and disassemble in hot cell. And then, according to the design requirements the non-instrumented DUPIC capsule was successfully manufactured. Also, the manufacturing technologies of the non-instrumented capsule for irradiating the nuclear fuel in HANARO were established, and the basic technology for the development of the instrumented capsule technology was accumulated. This report describes the manufacturing of the non-instrumented capsule for simulated DUPIC fuel. And, this report will be based to develope the instrumented capsule, which will be utilized to irradiate the nuclear fuel in HANARO. 26 refs., 4 figs. (Author)

  3. Conceptual design of experimental LFR fuel element for testing in TRIGA reactor, ACPR zone

    International Nuclear Information System (INIS)

    Nastase, D.; Olteanu, G.; Ioan, M.; Pauna, E.

    2013-01-01

    In the pulsed area of the TRIGA reactor (ACPR zone), the irradiation tests called ''rapid insertions of reactivity on different types of nuclear fuel elements'' are usually realized. During these tests, in the fuel element high powers for a relatively short period of time (about few milliseconds) are generated. The generated heat in fuel pellets raise their central temperature to values over 100 deg C. The conceptual design of an experimental fuel element proposed to be developed and presented in this paper must fulfill a couple of requirements, as follows: to ensure full compatibility with irradiation device sample holder (compatibility is achieved through reduced length of the fuel stack pellets - this way assures a flow flattening on the entire length of the fuel element); to be compatible with the project of irradiated fuel bundle in Lead cooled Fast Reactors (LFR). (authors)

  4. Review of FRAP-T4 performance based on fuel behavior tests conducted in the PBF

    International Nuclear Information System (INIS)

    Charyulu, M.K.

    1979-09-01

    The ability of the Fuel Rod Analysis Program - Transient (FRAP-T), a computer code developed at the Idaho National Engineering Laboratory to calculate fuel rod behavior during transient experiments conducted in the Power Burst Facility, is discussed. Fuel rod behavior calculations are compared with data from tests performed under postulated RIA, LOCA, and PCM accident conditions. Physical phenomena, rod damage, and damage mechanisms observed during the tests and not presently incorporated into the FRAP-T code are identified

  5. Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1990-06-01

    The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85 degree C and 25 degree C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs

  6. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    1983-08-01

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  7. Dynamical Capillary Rise Photonic Sensor for Testing of Diesel and Biodiesel Fuel

    Directory of Open Access Journals (Sweden)

    Michal BORECKI

    2015-10-01

    Full Text Available There are many fuel quality standards introduced by national organizations and fuel producers. Usual techniques for measuring fuel parameters like cetane number, cetane index, fraction composition, viscosity, density, and flash point, require relatively complex and expensive laboratory equipment. On the fuel user side, fast and low cost sensing of useful state of biodiesel fuel is important. The main parameters of diesel fuel compatibility are: density, viscosity and surface tension. These three parameters define indirectly the quality of the fuel atomization process and the injected portion of energy that affect the quality of the fuel. In the presented paper the purposefulness of fuel testing using measurements of separable parameters is discussed. On this base, a sensor which enables the examination of relation of the mentioned parameters in one arrangement is proposed, analyzed and tested. The sensor uses the dynamic capillary rise method with photonic multichannel data reading in an inclined capillary. The principle of the sensor’s operation, the construction of the sensor head, and the experimental results are presented. The capillary is a disposable element. The sensor testing was performed with freshly prepared biodiesel fuels, and fuels stored for 2 years. We conclude that the proposed construction may be in future the base of low cost commercially marketable instruments for basic fuel classification: fit for use or not. That classification includes initial fuel composition and fuel parameters change during storage. Therefore, the proposed sensor is intended to use in fuel buying/selling point rather than used as part of a diesel engine automated system.

  8. Allocation of multiple, widely spread oil spills associated with one polluter : GC-MS fingerprinting and diagnostic ratios of spilled oil and oiled seabirds

    International Nuclear Information System (INIS)

    Hansen, A.B.; Avnskjold, J.

    2005-01-01

    In January 2005, a Cypriot cargo ship leaked about 5 tons of heavy fuel bunker oil in Kerteminde Bay in the Great Belt, Denmark. The ship was stopped to inspect and collect oil samples from its 2 damaged tanks for forensic oil spill identification by gas chromatography and mass spectrometry. Two weeks following the accident, a series of waterborne and stranded oil spills showed up in the Great Belt area, north and south of the vessel's route. Thousands of oiled seabirds on small islands and coastlines were affected. The Danish Coast Guard suspected that the vessel might be responsible for the observed spills. More than 50 oil samples were collected and sent for forensic analysis at the National Environmental Research Institute. Both waterborne and stranded spill samples showed an almost perfect match of diagnostic ratios and chromatograph with the potential responsible party (PRP) bunker. The spill samples therefore matched the reference oil and were allocated to the spill associated with the Cypriot cargo ship. One sample deviated significantly from the other samples and was not allocated to the ship's accidental spill. Oil samples collected from oiled seabirds showed larger variations between diagnostic ratios and the reference bunker oils. The variations can be attributed to weathering and biodegradation, but also to contamination by non-petrogenic material. It was concluded that the oiled seabirds represented non-match samples that cannot be allocated to the oil spill associated with the Cypriot cargo ship. 14 refs., 3 tabs., 4 figs

  9. Improving the AGR fuel testing power density profile versus irradiation-time in the advanced test reactor

    International Nuclear Information System (INIS)

    Chang, Gray S.; Lillo, Misti A.; Maki, John T.; Petti, David A.

    2009-01-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235 U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235 U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250degC throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235 U

  10. Solid-fuel cook stoves: Fuel efficiency and emissions testing--Austin

    Science.gov (United States)

    The World Health Organization estimates that approximately 1.6 million people prematurely die each year due to exposure to air pollutants from burning solid fuels for residential cooking and heating (WHO, 2010). Residential solid-fuel use accounts for approximately 25 percent of ...

  11. Bioremediation of oil spills

    International Nuclear Information System (INIS)

    Webb, M.

    1992-01-01

    For some years now UK and European oil spill response agencies, together with oil companies having an exploration or production interest in the European area, have been developing interest in the possible use of bioremediation techniques in combatting oil spills. The interest has accelerated in the aftermath of Exxon Valdez but there is significant scepticism over the actual value of the technique. The promise of increased rates of oil degradation, using bacteria or nutrients, does not yet appear to have been properly validated and there is concern over possible knock-on environmental effects. In consequence the response agencies are reluctant to bring the technique into their current combat armory. Some of the questions raised are: What efficacious techniques are available and how were they proven? On what type of oils can they be used? What is the scope for their use (at sea, type of coastline, temperature limitations, etc.)? What are the short and long term effects? Does bioremediation really work and offer a potential tool for oil spill clean-up? How do cleaning rates compare with natural recovery? There are many others. The view of the European Commission is that there should be a coordinated effort to answer these questions, but that effort should be properly targeted. I concur strongly with this view. The tasks are too large and varied for piecemeal attention. The European Commission wishes to initiate appropriate coordinated work, directed at the needs of European nations but which will subsequently inform the international response community through the International Maritime Organization and its Oil Pollution Preparedness and Response Cooperation initiative

  12. Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

    International Nuclear Information System (INIS)

    Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

    1986-01-01

    TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

  13. Marine oil spill response organizations

    International Nuclear Information System (INIS)

    Hendry, C.

    1997-01-01

    The obligations under the law relative to the prevention of marine oil spills and the type of emergency plans needed to mitigate any adverse effects caused by a marine oil spill were discussed. The organizational structure, spill response resources and operational management capabilities of Canada's newly created Response Organizations (ROs) were described. The overall range of oil spill response services that the RO provides to the domestic oil handling, oil transportation and the international shipping industries were reviewed. Amendments to the Canada Shipping Act which require that certain ships and oil handling facilities take oil spill preparedness and response measures, including having an arrangement with an RO certified by the Canadian Coast Guard, were outlined. Canadians now benefit from five ROs established to provide coast-to-coast oil spill response coverage. These include the Western Canada Marine Response Corporation, the Canadian Marine Response Management Corporation, the Great Lakes Response Corporation, the Eastern Canada Response Corporation and the Atlantic Emergency Response Team Ltd. ROs have the expertise necessary to organize and manage marine oil spill response services. They can provide equipment, personnel and operational management for the containment, recovery and cleanup of oil spilled on water

  14. Oil spill clean up

    International Nuclear Information System (INIS)

    Claxton, L.D.; Houk, V.S.; Williams, R.; Kremer, F.

    1991-01-01

    Due to the consideration of bioremediation for oil spills, it is important to understand the ecological and human health implications of bioremediation efforts. During biodegradation, the toxicity of the polluting material may actually increase upon the conversion of non-toxic constituents to toxic species. Also, toxic compounds refractory to biological degradation may compromise the effectiveness of the treatment technique. In the study, the Salmonella mutagenicity assay showed that both the Prudhoe Bay crude oil and its weathered counterpart collected from oil-impacted water were weakly mutagenic. Results also showed that the mutagenic components were depleted at a faster rate than the overall content of organic material

  15. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    Sudjatmi KA

    2013-01-01

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

  16. Acute aquatic toxicity of heavy fuel oils. Summary of relevant test data

    Energy Technology Data Exchange (ETDEWEB)

    Comber, M.I.H.; Den Haan, K.; Djemel, N.; Eadsforth, C.V.; King, D.; Parkerton, T.; Paumen, M.L.; Dmytrasz, B.

    2011-12-15

    This report describes the experimental procedures and results obtained in acute ecotoxicity tests on several heavy fuel oil (HFO) samples. Water accommodated fractions (WAFs) of these samples were tested for toxicity to the rainbow trout (Oncorhynchus mykiss), the crustacean zooplankter (Daphnia magna) and green algae (Selenastrum capricornutum). These results assist in determining the environmental hazard from heavy fuel oil.

  17. Acute aquatic toxicity of heavy fuel oils. Summary of relevant test data

    International Nuclear Information System (INIS)

    Comber, M.I.H.; Den Haan, K.; Djemel, N.; Eadsforth, C.V.; King, D.; Parkerton, T.; Paumen, M.L.; Dmytrasz, B.

    2011-12-01

    This report describes the experimental procedures and results obtained in acute ecotoxicity tests on several heavy fuel oil (HFO) samples. Water accommodated fractions (WAFs) of these samples were tested for toxicity to the rainbow trout (Oncorhynchus mykiss), the crustacean zooplankter (Daphnia magna) and green algae (Selenastrum capricornutum). These results assist in determining the environmental hazard from heavy fuel oil.

  18. Field testing at the Climax Stock on the Nevada Test Site: spent fuel test and radionuclide migration experiments

    International Nuclear Information System (INIS)

    Ballou, L.B.; Isherwood, D.J.; Patrick, W.C.

    1982-01-01

    Two field tests in the Climax Stock are being conducted. The Climax Stock, a granitic instrusive, has been administratively excluded from consideration as a full-scale repository site. However, it provides a readily available facility for field testing with high-level radioactive materials at a depth (420 m) approaching that of a repository. The major test activity in the 1980 fiscal year has been initiation of the Spent Fuel Test-Climax (SFT-C). This test, which was authorized in June 1978, is designed to evaluate the generic feasibility of geologic storage and retrievability of commercial power reactor spent fuel assemblies in a granitic medium. In addition, the test is configured and instrumented to provide thermal and thermomechanical response data that will be relevant to the design of a repository in hard crystalline rock. The other field activity in the Climax Stock is a radionuclide migration test. It combines a series of field and laboratory migration experiments with the use of existing hydrologic models for pretest predictions and data interpretation. Goals of this project are to develop: (1) field measurement techniques for radionuclide migration studies in a hydrologic regime where the controlling mechanism is fracture permeability; (2) field test data on radionuclide migration; and (3) a comparison of laboratory- and field-measured retardation factors. This radionuclide migration test, which was authorized in the middle of the 1980 fiscal year, is in the preliminary design phase. The detailed program plan was prepared and subjected to formal peer review in August. In September/October researchers conducted preliminary flow tests with water in selected near-vertical fractures intersected by small horizontal boreholes. These tests were needed to establish the range of pressures, flow rates, and other operating parameters to be used in conducting the nuclide migration tests. 21 references, 14 figures, 1 table

  19. NASA fuel cell applications for space: Endurance test results on alkaline fuel cell electrolyzer components

    International Nuclear Information System (INIS)

    Sheibley, D.W.

    1984-01-01

    Fuel cells continue to play a major role in manned spacecraft power generation. The Gemini and Apollo programs used fuel cell power plants as the primary source of mission electrical power, with batteries as the backup. The current NASA use for fuel cells is in the Orbiter program. Here, low temperature alkaline fuel cells provide all of the on-board power with no backup power source. Three power plants per shipset are utilized; the original power plant contained 32-cell substacks connected in parallel. For extended life and better power performance, each power plant now contains three 32-cell substacks connected in parallel. One of the possible future applications for fuel cells will be for the proposed manned Space Station in low earth orbit (LEO)(1, 2, 3). By integrating a water electrolysis capability with a fuel cell (a regenerative fuel cell system), a multikilowatt energy storage capability ranging from 35 kW to 250 kW can be achieved. Previous development work on fuel cell and electrolysis systems would tend to minimize the development cost of this energy storage system. Trade studies supporting initial Space Station concept development clearly show regenerative fuel cell (RFC) storage to be superior to nickel-cadmium and nickel-hydrogen batteries with regard to subsystem weight, flexibility in design, and integration with other spacecraft systems when compared for an initial station power level ranging from 60 kW to 75 kW. The possibility of scavenging residual O 2 and H 2 from the Shuttle external tank for use in fuel cells for producing power also exists

  20. Technical concept for test of geologic storage of spent reactor fuel in the Climax granite, Nevada Test Site

    International Nuclear Information System (INIS)

    Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

    1979-01-01

    The Spent Fuel Test in the Climax granite at the Nevada Test Site is a generic test in which spent fuel assemblies from an operating commercial nuclear reactor are emplaced at, and retrieved from, a plausible waste repository depth in a typical granite. Eleven canisters of spent fuel are emplaced in a storage drift 420 m below the surface along with six electrical simulator canisters. Two adjacent drifts contain electrical heaters which are operated so as to simulate the initial five years of the temperature-stress-displacement fields of a large repository. The site is described, and the pre-operational measurement program and characteristics of the spent fuel are given. Both thermal and mechanical response calculations are summarized. The field instrumentation and data acquisition systems are described, as well as the system for handling the spent fuel

  1. Fabrication and testing of uranium nitride fuel for space power reactors

    Science.gov (United States)

    Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

    1988-02-01

    Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

  2. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  3. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

    2013-01-01

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

  4. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  5. Mitigating oil spills in the water column

    International Nuclear Information System (INIS)

    Barry, Edward; Libera, Joseph A.; Mane, Anil University; Avila, Jason R.; DeVitis, David

    2017-01-01

    The scale and scope of uncontrolled oil spills can be devastating. Diverse marine environments and fragile ecologies are some of the most susceptible to the many ill effects, while the economic costs can be crippling. A notoriously difficult challenge with no known technological solution is the successful removal of oil dispersed in the water column. Here, we address this problem through cheap and reusable oil sorbents based on the chemical modification of polymer foams. Interfacial chemistry was optimized and subsequently tested in a simulated marine environment at the National Oil Spill Response Research & Renewable Energy Test Facility, Ohmsett. We find favorable performance for surface oil mitigation and, for the first time, demonstrate the advanced sorbent's efficiency and efficacy at pilot scale in extraction of crude oil and refined petroleum products dispersed in the water column. As a result, this is a potentially disruptive technology, opening a new field of environmental science focused on sub-surface pollutant sequestration.

  6. National Fuel Cell Bus Program : Accelerated Testing Report, AC Transit

    Science.gov (United States)

    2009-01-01

    This is an evaluation of hydrogen fuel cell transit buses operating at AC Transit in revenue service since March 20, 2006 compared to similar diesel buses operating from the same depot. This evaluation report includes results from November 2007 throu...

  7. Test system design for Hardware-in-Loop evaluation of PEM fuel cells and auxiliaries

    Energy Technology Data Exchange (ETDEWEB)

    Randolf, Guenter; Moore, Robert M. [Hawaii Natural Energy Institute, University of Hawaii, Honolulu, HI (United States)

    2006-07-14

    In order to evaluate the dynamic behavior of proton exchange membrane (PEM) fuel cells and their auxiliaries, the dynamic capability of the test system must exceed the dynamics of the fastest component within the fuel cell or auxiliary component under test. This criterion is even more critical when a simulated component of the fuel cell system (e.g., the fuel cell stack) is replaced by hardware and Hardware-in-Loop (HiL) methodology is employed. This paper describes the design of a very fast dynamic test system for fuel cell transient research and HiL evaluation. The integration of the real time target (which runs the simulation), the test stand PC (that controls the operation of the test stand), and the programmable logic controller (PLC), for safety and low-level control tasks, into one single integrated unit is successfully completed. (author)

  8. Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walters, L. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO2 and UO2-PuO2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.

  9. Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio

    2000-11-01

    Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)