46 CFR 119.470 - Ventilation of spaces containing diesel fuel tanks.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Ventilation of spaces containing diesel fuel tanks. 119... fuel tanks. (a) Unless provided with ventilation that complies with § 119.465 of this part, a space containing a diesel fuel tank and no machinery must meet one of the following requirements: (1) A space of 14...

Development of fast reactor metal fuels containing minor actinides

International Nuclear Information System (INIS)

Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

2011-01-01

Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

Measuring method for effective neutron multiplication factor upon containing irradiated fuel assembly

International Nuclear Information System (INIS)

Ueda, Makoto; Mitsuhashi, Ishi; Sasaki, Tomoharu.

1993-01-01

A portion of irradiated fuel assemblies at a place where a reactivity effect is high, that is, at a place where neutron importance is high is replaced with standard fuel assemblies having a known composition to measure neutron fluxes at each of the places. An effective composition at the periphery of the standard fuel assemblies is determined by utilizing a calibration curve determined separately based on the composition and neutron flux values of the standard assemblies. By using the calibration curve determined separately based on this composition and the known composition of the standard fuel assemblies, an effective neutron multiplication factor for the fuel containing portion containing the irradiated fuel assemblies is recognized. Then, subcriticality is ensured and critical safety upon containing the fuel assemblies can be secured quantitatively. (N.H.)

Multi-purpose container technologies for spent fuel management

International Nuclear Information System (INIS)

2000-12-01

The management of spent nuclear fuel is an integral part of the nuclear fuel cycle. Spent fuel management resides in the back end of the fuel cycle, and is not revenue producing as electric power generation is. It instead results in a cost associated power generation. It is a major consideration in the nuclear power industry today. Because technologies, needs and circumstances vary from country to country, there is no single, standardized approach to spent fuel management. The projected cumulative amount of spent fuel generated worldwide by 2010 will be 330 000 t HM. When reprocessing is accounted for, that amount is likely to be reduced to 215 000 t HM, which is still more than twice as much as the amount now in storage. Considering the limited capacity of at-reactor (AR) storage, various technologies are being developed for increasing storage capacities. At present, many countries are developing away-from-reactor (AFR) storage in the form of pool storage or as dry storage. Further these AFR storage systems may be at-reactor sites or away-from-reactor sites (e.g. centrally located interim storage facilities, serving several reactors). The dry storage technologies being developed are varied and include vaults, horizontal concrete modules, concrete casks, and metal casks. The review of the interim storage plans of several countries indicates that the newest approaches being pursued for spent fuel management use dual-purpose and multi-purpose containers. These containers are envisaged to hold several spent fuel assemblies, and be part of the transport, storage, and possibly geological disposal systems of an integrated spent fuel management system

PRODUCTION OF NEW BIOMASS/WASTE-CONTAINING SOLID FUELS

Energy Technology Data Exchange (ETDEWEB)

David J. Akers; Glenn A. Shirey; Zalman Zitron; Charles Q. Maney

2001-04-20

CQ Inc. and its team members (ALSTOM Power Inc., Bliss Industries, McFadden Machine Company, and industry advisors from coal-burning utilities, equipment manufacturers, and the pellet fuels industry) addressed the objectives of the Department of Energy and industry to produce economical, new solid fuels from coal, biomass, and waste materials that reduce emissions from coal-fired boilers. This project builds on the team's commercial experience in composite fuels for energy production. The electric utility industry is interested in the use of biomass and wastes as fuel to reduce both emissions and fuel costs. In addition to these benefits, utilities also recognize the business advantage of consuming the waste byproducts of customers both to retain customers and to improve the public image of the industry. Unfortunately, biomass and waste byproducts can be troublesome fuels because of low bulk density, high moisture content, variable composition, handling and feeding problems, and inadequate information about combustion and emissions characteristics. Current methods of co-firing biomass and wastes either use a separate fuel receiving, storage, and boiler feed system, or mass burn the biomass by simply mixing it with coal on the storage pile. For biomass or biomass-containing composite fuels to be extensively used in the U.S., especially in the steam market, a lower cost method of producing these fuels must be developed that includes both moisture reduction and pelletization or agglomeration for necessary fuel density and ease of handling. Further, this method of fuel production must be applicable to a variety of combinations of biomass, wastes, and coal; economically competitive with current fuels; and provide environmental benefits compared with coal. Notable accomplishments from the work performed in Phase I of this project include the development of three standard fuel formulations from mixtures of coal fines, biomass, and waste materials that can be used in

Hybrid laser arc welding of a used fuel container

Energy Technology Data Exchange (ETDEWEB)

Boyle, C., E-mail: cboyle@nwmo.ca [Nuclear Waste Management Organization, Toronto, ON (Canada); Martel, P. [Novika Solutions, La Pocatiere, QC (Canada)

2015-07-01

The Nuclear Waste Management Organization (NWMO) has designed a novel Used Fuel Container (UFC) optimized for CANDU used nuclear fuel. The Mark II container is constructed of nuclear grade pipe for the body and capped with hemi-spherical heads. The head-to-shell joint fit-up features an integral backing designed for external pressure, eliminating the need for a full penetration closure weld. The NWMO and Novika Solutions have developed a partial penetration, single pass Hybrid Laser Arc Weld (HLAW) closure welding process requiring no post-weld heat treatment. This paper will discuss the joint design, HLAW process, associated welding equipment, and prototype container fabrication. (author)

Hybrid laser arc welding of a used fuel container

Energy Technology Data Exchange (ETDEWEB)

Boyle, C. [Nuclear Waste Management Organization (NWMO), Toronto, Ontario (Canada); Martel, P. [Novika Solutions, La Pocatiere, Quebec (Canada)

2015-09-15

The Nuclear Waste Management Organization (NWMO) has designed a novel Used Fuel Container (UFC) optimized for CANDU used nuclear fuel. The Mark II container is constructed of nuclear grade pipe for the body and capped with hemi-spherical heads. The head-to-shell joint fit-up features an integral backing designed for external pressure, eliminating the need for a full penetration closure weld. The NWMO and Novika Solutions have developed a partial penetration, single pass Hybrid Laser Axe Weld (HLAW) closure welding process requiring no post-weld heat treatment. This paper will discuss the joint design, HLAW process, associated welding equipment, and prototype container fabrication. (author)

Researches of real observation geometry in monitoring fuel-containing materials' subcriticality

International Nuclear Information System (INIS)

Vysotskij, E.D.; Shevchenko, V.G.; Shevchenko, M.V.

2002-01-01

The effectiveness of fuel-containing materials monitoring is discussed in the part related to the feasibilities of researches and realization of optimal geometry (detectors - source) of survey of neutron activity dynamics in nuclearly hazardous areas with clusters of fuel-containing materials concentrated in the premises 305/2

Two-dimensional radiation shielding optimization analysis of spent fuel transport container

International Nuclear Information System (INIS)

Tian Yingnan; Chen Yixue; Yang Shouhai

2013-01-01

The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

Waste water plant for radioactive polluted media, for example faeces

International Nuclear Information System (INIS)

Galler, L.

1985-01-01

In order to provide simple transport of radioactive polluted wastewater to a decay container using simple structural means, where aerobic decay is to take place during storage in the decay container, it is proposed that a sub-pressure source should be connected to the place of origin via a pressure-tight and closed decay container and that the gasous fluid extracted by the subpressure source should be returned to the decay container. (orig./PW) [de

Pressurized solid oxide fuel cell integral air accumular containment

Science.gov (United States)

Gillett, James E.; Zafred, Paolo R.; Basel, Richard A.

2004-02-10

A fuel cell generator apparatus contains at least one fuel cell subassembly module in a module housing, where the housing is surrounded by a pressure vessel such that there is an air accumulator space, where the apparatus is associated with an air compressor of a turbine/generator/air compressor system, where pressurized air from the compressor passes into the space and occupies the space and then flows to the fuel cells in the subassembly module, where the air accumulation space provides an accumulator to control any unreacted fuel gas that might flow from the module.

Thermal conditions and functional requirements for molten fuel containment

International Nuclear Information System (INIS)

Kang, C.S.; Torri, A.

1980-05-01

This paper discusses the configuration and functional requirements for the molten fuel containment system (MFCS) in the GCFR demonstration plant design. Meltdown conditions following a loss of shutdown cooling (LOSC) accident were studied to define the core debris volume for a realistic meltdown case. Materials and thicknesses of the molten fuel container were defined. Stainless steel was chosen as the sacrificial material and magnesium oxide was chosen as the crucible material. Thermal conditions for an expected quasi-steady state were analyzed. Highlights of the functional requirements which directly affect the MFCS design are discussed

Palliative effects of H2 on SOFCs operating with carbon containing fuels

Science.gov (United States)

Reeping, Kyle W.; Bohn, Jessie M.; Walker, Robert A.

2017-12-01

Chlorine can accelerate degradation of solid oxide fuel cell (SOFC) Ni-based anodes operating on carbon containing fuels through several different mechanisms. However, supplementing the fuel with a small percentage of excess molecular hydrogen effectively masks the degradation to the catalytic activity of the Ni and carbon fuel cracking reaction reactions. Experiments described in this work explore the chemistry behind the "palliative" effect of hydrogen on SOFCs operating with chlorine-contaminated, carbon-containing fuels using a suite of independent, complementary techniques. Operando Raman spectroscopy is used to monitor carbon accumulation and, by inference, Ni catalytic activity while electrochemical techniques including electrochemical impedance spectroscopy and voltammetry are used to monitor overall cell performance. Briefly, hydrogen not only completely hides degradation observed with chlorine-contaminated carbon-containing fuels, but also actively removes adsorbed chlorine from the surface of the Ni, allowing for the methane cracking reaction to continue, albeit at a slower rate. When hydrogen is removed from the fuel stream the cell fails immediately due to chlorine occupation of methane/biogas reaction sites.

Casting of metallic fuel containing minor actinide additions

International Nuclear Information System (INIS)

Trybus, C.L.; Henslee, S.P.; Sanecki, J.E.

1992-01-01

A significant attribute of the Integral Fast Reactor (IFR) concept is the transmutation of long-lived minor actinide fission products. These isotopes require isolation for thousands of years, and if they could be removed from the waste, disposal problems would be reduced. The IFR utilizes pyroprocessing of metallic fuel to separate auranium, plutonium, and the minor actinides from nonfissionable constituents. These materials are reintroduced into the fuel and reirradiated. Spent IFR fuel is expected to contain low levels of americium, neptunium, and curium because the hard neutron spectrum should transmute these isotopes as they are produced. This opens the possibility of using an IFR to trnasmute minor actinide waste from conventional light water reactors (LWRs). A standard IFR fuel is based on the alloy U-20% Pu-10% Zr (in weight percent). A metallic fuel system eases the requirements for reprocessing methods and enables the minor actinide metals to be incorporated into the fuel with simple modifications to the basic fuel casting process. In this paper, the authors report the initial casting experience with minor actinide element addition to an IFR U-Pu-Zr metallic fuel

Estimates of Particulate Mass for an MCO Containing Mark 1A Fuel

International Nuclear Information System (INIS)

WYMAN, H.A.

1999-01-01

High, best estimate, and low values are given for particulate inventories within an MCO basket containing freshly cleaned Mark 1A fuel. The findings are compared with the estimates of particulate inventories for an MCO basket containing freshly cleaned Mark IV fuel

CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

Energy Technology Data Exchange (ETDEWEB)

Vinson, D.

2010-07-11

Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and

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

International Nuclear Information System (INIS)

Ferreira, W.J.; Luz, M.

1985-01-01

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

ETR fuel element shipping container addendum to PR-T-79-011 (TR-466). Internal technical report

International Nuclear Information System (INIS)

Smith, M.C.

1979-01-01

In July, 1979, EG and G Idaho, Inc. was requested to evaluate the ETR Fuel Element Shipping Container for compliance with existing transport regulations, in order to ship GETR fuel elements from Vallecitos, California to the INEL. Technical report PR-T-79-011 (TR-466), ATR Fuel Element Shipping Container Safety Analysis, was used as a basis for this evaluation. The safety analysis contained in technical report PR-T-79-011 (TR-466) was performed utilizing the ATR, ETR, MTR, and SPERT shipping containers. The report determined the ETR Fuel Element Shipping Container does comply with the existing transport regulations for a Type A quantity, Fissile Class I shipping container. The ETR and GETR fuel elements are essentially identical in physical size, construction, and fissile material content, the analysis documented in this report has determined the shipment of GETR fuel elements in the ETR shipping container to be safe and pose no threat to the public health and safety

Assessment of energy performance and air pollutant emissions in a diesel engine generator fueled with water-containing ethanol-biodiesel-diesel blend of fuels

International Nuclear Information System (INIS)

Lee, Wen-Jhy; Liu, Yi-Cheng; Mwangi, Francis Kimani; Chen, Wei-Hsin; Lin, Sheng-Lun; Fukushima, Yasuhiro; Liao, Chao-Ning; Wang, Lin-Chi

2011-01-01

Biomass based oxygenated fuels have been identified as possible replacement of fossil fuel due to pollutant emission reduction and decrease in over-reliance on fossil fuel energy. In this study, 4 v% water-containing ethanol was mixed with (65-90%) diesel using (5-30%) biodiesel (BD) and 1 v% butanol as stabilizer and co-solvent respectively. The fuels were tested against those of biodiesel-diesel fuel blends to investigate the effect of addition of water-containing ethanol for their energy efficiencies and pollutant emissions in a diesel-fueled engine generator. Experimental results indicated that the fuel blend mix containing 4 v% of water-containing ethanol, 1 v% butanol and 5-30 v% of biodiesel yielded stable blends after 30 days standing. BD1041 blend of fuel, which composed of 10 v% biodiesel, 4 v% of water-containing ethanol and 1 v% butanol demonstrated -0.45 to 1.6% increase in brake-specific fuel consumption (BSFC, mL kW -1 h -1 ) as compared to conventional diesel. The better engine performance of BD1041 was as a result of complete combustion, and lower reaction temperature based on the water cooling effect, which reduced emissions to 2.8-6.0% for NO x , 12.6-23.7% particulate matter (PM), 20.4-23.8% total polycyclic aromatic hydrocarbons (PAHs), and 30.8-42.9% total BaPeq between idle mode and 3.2 kW power output of the diesel engine generator. The study indicated that blending diesel with water-containing ethanol could achieve the goal of more green sustainability. -- Highlights: → Water-containing ethanol was mixed with diesel using biodiesel and butanol as stabilizer and co-solvent, respectively. → Fuel blends with 4 v% water-containing ethanol, 1 v% butanol, 5-30 v% biodiesel and conventional diesel yielded a stable blended fuel after more than 30 days. → Due to more complete combustion and water quench effect, target fuel BD1041 was gave good energy performance and significant reduction of PM, NO x , total PAH and total BaPeq emissions. �

Fault tree analysis of the manufacturing process of nuclear fuel containers

International Nuclear Information System (INIS)

Liao Weixian; Men Dechun; Sui Yuxue

1998-08-01

The nuclear fuel container consists of barrel body, bottom, cover, locking ring, rubber seal ring, and so on. It should be kept sealed in transportation and storage, so keeps the fuel contained from leakage. Its manufacturing process includes blanking, forming, seam welding, assembling, derusting and painting. The seam welding and assembling of barrel body and bottom are two key procedures, and the slope grinding, barrel body flaring and deep drawing of the bottom are important procedures. Faults in the manufacturing process of the nuclear fuel containers are investigated in details as for its quality requirements. A fault tree is established with products being unqualified as the top event. Five causes resulting in process faults are classified and analysed, and some measures are suggested for controlling different failures in manufacturing. More research work should be conducted in rules how to set up fault trees for manufacturing process

Pyrometallurgical separation processes of radionuclides contained in the irradiated nuclear fuel

International Nuclear Information System (INIS)

De Cordoba, Guadalupe; Caravaca, Concha; Quinones, Javier; Gonzalez de la Huebra, Angel

2005-01-01

Faced with the new options for the high level waste management, the ''Partitioning and Transmutation (P and T)'' of the radio nuclides contained in the irradiated nuclear fuel appear as a promising option from different points of view, such as environmental risk, radiotoxic inventory reduction, economic, etc.. The present work is part of a research project called ''PYROREP'' of the 5th FWP of the EU that studied the feasibility of the actinide separation from the rest of fission products contained in the irradiated nuclear fuel by pyrometallurgical processes with the aim of their transmutation. In order to design these processes it is necessary to determine basic thermodynamic and kinetic data of the radionuclides contained in the nuclear fuel in molten salt media. The electrochemical study of uranium, samarium and molybdenum in the eutectic melt LiCl - KCl has been performed at a tungsten electrode in the temperature range of 450 - 600 deg C in order to obtain these basic properties. (Author)

Application of fuel cell for pyrite and heavy metal containing mining waste

Science.gov (United States)

Keum, H.; Ju, W. J.; Jho, E. H.; Nam, K.

2015-12-01

Once pyrite and heavy metal containing mining waste reacts with water and air it produces acid mine drainage (AMD) and leads to the other environmental problems such as contamination of surrounding soils. Pyrite is the major source of AMD and it can be controlled using a biological-electrochemical dissolution method. By enhancing the dissolution of pyrite using fuel cell technology, not only mining waste be beneficially utilized but also be treated at the same time by. As pyrite-containing mining waste is oxidized in the anode of the fuel cell, electrons and protons are generated, and electrons moves through an external load to cathode reducing oxygen to water while protons migrate to cathode through a proton exchange membrane. Iron-oxidizing bacteria such as Acidithiobacillus ferrooxidans, which can utilize Fe as an electron donor promotes pyrite dissolution and hence enhances electrochemical dissolution of pyrite from mining waste. In this study mining waste from a zinc mine in Korea containing 17 wt% pyrite and 9% As was utilized as a fuel for the fuel cell inoculated with A. ferrooxidans. Electrochemically dissolved As content and chemically dissolved As content was compared. With the initial pH of 3.5 at 23℃, the dissolved As concentration increased (from 4.0 to 13 mg/L after 20 d) in the fuel cell, while it kept decreased in the chemical reactor (from 12 to 0.43 mg/L after 20 d). The fuel cell produced 0.09 V of open circuit voltage with the maximum power density of 0.84 mW/m2. Dissolution of As from mining waste was enhanced through electrochemical reaction. Application of fuel cell technology is a novel treatment method for pyrite and heavy metals containing mining waste, and this method is beneficial for mining environment as well as local community of mining areas.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Shale oil combustion

International Nuclear Information System (INIS)

Al-dabbas, M.A.

1992-05-01

A 'coutant' carbon steel combustion chamber cooled by water jacket was conslructed to burn diesel fuel and mixlure of shale oil and diesel fuels. During experimental work nir fuel ratio was determined, temperaturces were measured using Chromel/ Almel thermocouple, finally the gasous combustion product analysis was carricd out using gas chromatograph technique. The constructed combustion chamber was operating salisfactory for several hours of continous work. According to the measurements it was found that: the flame temperature of a mixture of diesel and shale oil fuels was greater than the flame temperature of diesel fuel. and the sulfer emissious of a mixture of diesel and shale oil fuels was higher than that of diesel fuel. Calculation indicated that the dry gas energy loss was very high and the incomplete combustion energy loss very small. (author). 23 refs., 35 figs

Shale oil combustion

Energy Technology Data Exchange (ETDEWEB)

Al-dabbas, M A

1992-05-01

A `coutant` carbon steel combustion chamber cooled by water jacket was conslructed to burn diesel fuel and mixlure of shale oil and diesel fuels. During experimental work nir fuel ratio was determined, temperaturces were measured using Chromel/ Almel thermocouple, finally the gasous combustion product analysis was carricd out using gas chromatograph technique. The constructed combustion chamber was operating salisfactory for several hours of continous work. According to the measurements it was found that: the flame temperature of a mixture of diesel and shale oil fuels was greater than the flame temperature of diesel fuel. and the sulfer emissious of a mixture of diesel and shale oil fuels was higher than that of diesel fuel. Calculation indicated that the dry gas energy loss was very high and the incomplete combustion energy loss very small. (author). 23 refs., 35 figs.

Thermal analysis of the IDENT 1578 fuel pin shipping container

International Nuclear Information System (INIS)

Ingham, J.G.

1980-01-01

The IDENT 1578 container, which is a 110-in. long 5.5-in. OD tube, is designed for shipping FFTF fuel elements in T-3 casks between HEDL, HFEF, and other laboratories. The thermal analysis was conducted to evaluate whether or not the container satisfies its thermal design criteria

Criticality safety evaluation for TWR-S fuel assembly transportation using TK-S16 containers

International Nuclear Information System (INIS)

Pesic, M.P.; Steljic, M.M.; Antic, D.P.

2002-01-01

Criticality safety issues, concerning transportation of fresh high-enriched uranium fuel elements (TWR-S fuel assembly type) with Russian containers TK-S16, are objects of study in this paper. Three-dimensional (3D) models of fuel element and container were made, based upon their well-known geometry and material structure. The way to pack fuel elements in a bundle inside of the container is proposed. Calculations were done by MCNP4B2 computer code. This Monte Carlo criticality code determined the effective multiplication factor from the cross-section data and specific geometry data. This evaluation demonstrated the subcriticality of a single package and an array of packages during normal conditions of transport and various hypothetical accident conditions. (author)

Container materials in environments of corroded spent nuclear fuel

Science.gov (United States)

Huang, F. H.

1996-07-01

Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

Development of a fresh plutonium fuel container for a prototype fast breeder reactor

International Nuclear Information System (INIS)

Ohtake, T.; Takahashi, S.; Mishima, T.; Kurakami, J.; Yamamoto, Y.; Ohuchi, Y.

1989-01-01

Japan gives a good deal of encouragement to development of a fast breeder reactor (which is considered as the most likely candidate for nuclear power generation) to secure long-term energy source. And, following an experimental fast breeder reactor Joyo, a prototype fast breeder reactor Monju is now under vigorous construction. Related to development of the prototype fast breeder reactor, it is necessary and important to develop transport container which is used for transporting fresh fuel assemblies from Plutonium Fuel Production Facility to the Monju power plant. Therefore, the container is now being developed by Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, shipment and vibration tests, handling performance tests, shielding performance tests and prototype container tests are executed with prototype containers fabricated according to a final design, in order to experimentally confirm soundness of transport container and its contents, and propriety of design technique. This paper describes the summary of general specifications and structures of this container and the summary of preliminary safety analysis of package

Shielding Performance Measurements of Spent Fuel Transportation Container

Directory of Open Access Journals (Sweden)

SUN Hong-chao

2015-11-01

Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

A method for determining the spent-fuel contribution to transport cask containment requirements

International Nuclear Information System (INIS)

Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Reardon, P.C.

1992-11-01

This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs

Pinhole Breaches in Spent Fuel Containers: Some Modeling Considerations

International Nuclear Information System (INIS)

Casella, Andrew M.; Loyalka, Sudarsham K.; Hanson, Brady D.

2006-01-01

This paper replaces PNNL-SA-48024 and incorporates the ANS reviewer's comments, including the change in the title. Numerical methods to solve the equations for gas diffusion through very small breaches in spent fuel containers are presented and compared with previous literature results

Corrosion assessment of dry fuel storage containers

Energy Technology Data Exchange (ETDEWEB)

Graves, C.E.

1994-09-01

The structural stability as a function of expected corrosion degradation of 75 dry fuel storage containers located in the 200 Area Low-Level Waste Burial Grounds was evaluated. These containers include 22 concrete burial containers, 13 55-gal (208-l) drums, and 40 Experimental Breeder Reactor II (EBR-II) transport/storage casks. All containers are buried beneath at least 48 in. of soil and a heavy plastic tarp with the exception of 35 of the EBR-II casks which are exposed to atmosphere. A literature review revealed that little general corrosion is expected and pitting corrosion of the carbon steel used as the exterior shell for all containers (with the exception of the concrete containers) will occur at a maximum rate of 3.5 mil/yr. Penetration from pitting of the exterior shell of the 208-l drums and EBR-II casks is calculated to occur after 18 and 71 years of burial, respectively. The internal construction beneath the shell would be expected to preclude containment breach, however, for the drums and casks. The estimates for structural failure of the external shells, large-scale shell deterioration due to corrosion, are considerably longer, 39 and 150 years respectively for the drums and casks. The concrete burial containers are expected to withstand a service life of 50 years.

Safety assessment of a dry storage container drop into irradiated fuel bays

International Nuclear Information System (INIS)

Parlatan, Y.; Oh, D.; Arguner, D.; Lei, Q.M.; Kulpa, T.; Bayoumi, M.H.

2004-01-01

In Pickering nuclear stations, Dry Storage Containers (DSCs) are employed to transfer used (irradiated) fuel from an irradiated fuel bay to a dry storage facility for interim storage. Each DSC is wet-loaded in the bay water with 4 fuel modules containing up to a total of 384 used fuel bundles that have been out of the reactor core for at least 10 years. Once the DSC is fully loaded, the crane in the bay raises the DSC for spray-wash such that the bottom of the DSC is never more than 2 m above the bay water surface. This paper presents a safety assessment of consequences of an unlikely event that a fully loaded DSC is accidentally dropped into an irradiated fuel bay from the highest possible elevation. Experiments and analyses performed elsewhere show that the DSC drop-generated shock waves will not threaten the structural integrity of an irradiated fuel bay. Therefore, this assessment only assesses the potential damage to the spent fuel bundles in the bay due to pressure transients generated by an accidental DSC drop. A bounding estimate approach has been used to calculate the upper limit of the pressure pulse and the resulting static and dynamic stresses on the fuel sheath. The bounding calculations and relevant experimental results demonstrate that an accidental drop of a fully loaded DSC into an irradiated fuel bay will not cause additional failures of the main fuel inventories stored in modules in the bay water, thus no consequential release of fission products into the bay water. (author)

International nuclear fuel cycle fact book. [Contains glossary

Energy Technology Data Exchange (ETDEWEB)

Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

1987-01-01

As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

Process and device for cooling of nuclear reactor fuel elements enclosed in a transport container

International Nuclear Information System (INIS)

Stiefel, M.

1986-01-01

In order to remove the post-decay heat of the fuel elements contained in them, transport containers for burnt-up fuel elements can be connected to a water cooling circuit. In order to avoid thermal shocks, a tenside forming foam and air are introduced into the cooling circuit before its entry into the transport container in the direction of flow. The tenside and air continue to be supplied until the temperature inside the transport container has fallen below the temperature at which the foam is destroyed. By adding tenside and air, a two phase mixture is produced, which foams greatly when it enters the transport container and which cools the fuel elements so as to protect them.(orig./HP) [de

Safety analysis report for packaging: the ORNL HFIR unirradiated fuel element shipping container

International Nuclear Information System (INIS)

Evans, J.H.; Boulet, J.A.M.; Eversole, R.E.

1977-11-01

The ORNL HFIR unirradiated fuel element shipping container was designed and fabricated at the Oak Ridge National Laboratory for the transport of HFIR unirradiated fuel elements. The container was evaluated analytically and experimentally to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported, and the evaluation is the subject of this report. Computational and test procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation demonstrate that the container is in compliance with the applicable regulations

Advanced containment research for the Canadian Nuclear Fuel Waste Management Program

International Nuclear Information System (INIS)

Onofrei, M.; Mathew, P.M.; McKay, P.; Hosaluk, L.J.; Oscarson, D.W.

1986-09-01

This document outlines the program on the development of advanced containment systems for the disposal of used fuel in a vault deep in plutonic rock. Possible advanced containment concepts, the strategy adopted in selecting potential container materials, and experimental programs currently underway or planned are presented. Most effort is currently directed toward developing long-term containment systems based on non-metallic materials and massive metal containers. The use of additional independent barriers to extend the lifetime of simple containment systems is also being evaluated. 58 refs

TECHNOLOGY FOR EFFICIENT USAGE OF HYDROCARBON-CONTAINING WASTE IN PRODUCTION OF MULTI-COMPONENT SOLID FUEL

Directory of Open Access Journals (Sweden)

B. M. Khroustalev

2016-01-01

Full Text Available The paper considers modern approaches to usage of hydrocarbon-containing waste as energy resources and presents description of investigations, statistic materials, analysis results on formation of hydrocarbon-containing waste in the Republic of Belarus. Main problems pertaining to usage of waste as a fuel and technologies for their application have been given in the paper. The paper describes main results of the investigations and a method for efficient application of viscous hydrocarbon-containing waste as an energy-packed component and a binding material while producing a solid fuel. A technological scheme, a prototype industrial unit which are necessary to realize a method for obtaining multi-component solid fuel are represented in the paper. A paper also provides a model of technological process with efficient sequence of technological operations and parameters of optimum component composition. Main factors exerting significant structure-formation influence in creation of structural composition of multi-component solid fuel have been presented in the paper. The paper gives a graphical representation of the principle for selection of mixture particles of various coarseness to form a solid fuel while using a briquetting method and comprising viscous hydrocarbon-containing waste. A dependence of dimensionless concentration g of emissions into atmosphere during burning of two-component solid fuel has been described in the paper. The paper analyzes an influence of the developed methodology for emission calculation of multi-component solid fuels and reveals a possibility to optimize the component composition in accordance with ecological function and individual peculiar features of fuel-burning equipment. Special features concerning storage and transportation, advantages and disadvantages, comparative characteristics, practical applicability of the developed multi-component solid fuel have been considered and presented in the paper. The paper

Assessing reliability and useful life of containers for disposal of irradiated fuel waste

International Nuclear Information System (INIS)

Doubt, G.

1984-06-01

Metal containers for fuel waste isolation are to be designed to last at least 500 years to provide a redundant barrier during the decay period of the high activity components of the waste. To meet the long-life requirement, containers must have a very low failure rate during the design mission, a low incidence of 'juvenile failures' due to undetected defects, and resistance to progressive deterioration from environmental processes. This paper summarizes studies to determine: (1) precedent for low failure rates and relevance to container longevity; (b) the likelihood of initial defects perforating the container before or shortly after emplacement, and estimates of material defect distribution; (c) the utility of reliability analysis techniques for estimating reliability and life of fuel waste containers; (d) other approaches to estimating container longevity and failure versus time distribution

Container for irradiated fuel

International Nuclear Information System (INIS)

Guy, R.

1978-01-01

The transport container for irradiated or used nuclear fuel is provided with an identical heat shield against fires on the top and bottom sides. Each heat shield consists of two inner nickel plates, whose contact surfaces are polished to a mirror finish and an outer plate of stainless steel. The nickel plate on the box is spot welded to it while the second nickel plate is spot welded to the steel plate. Both together are in turn welded so as to be leaktight to the edges of the box. For extreme heat effects and based on the different (bimetal) coefficients of expansion, the steel plate with the nickel plate attached to it bulges away from the box. The second nickel plate remains at the box, so that a subpressure space is formed with the mirror nickel surfaces. The heat radiation and heat conduction to the box are greatly reduced by this. (DG) [de

Nuclear analysis of the Chornobyl fuel containing masses with heterogeneous fuel distribution

International Nuclear Information System (INIS)

Turski, R. B.

1998-01-01

Although significant data has been obtained on the condition and composition of the fuel containing masses (FCM) located in the concrete chambers under the Chernobyl Unit 4 reactor cavity, there is still uncertainty regarding the possible recriticality of this material. The high radiation levels make access extremely difficult, and most of the samples are from the FCM surface regions. There is little information on the interior regions of the FCM, and one cannot assume with confidence that the surface measurements are representative of the interior regions. Therefore, reasonable assumptions on the key parameters such as fuel concentration, the concentrations of impurities and neutron poisons (especially boron), the void fraction of the FCM due to its known porosity, and the degrees of fuel heterogeneity, are necessary to evaluate the possibility of recriticality. The void fraction is important since it introduces the possibility of water moderator being distributed throughout the FCM. Calculations indicate that the addition of 10 to 30 volume percent (v/o) water to the FCM has a significant impact on the calculated reactivity of the FCM. Therefore, water addition must be considered carefully. The other possible moderators are graphite and silicone dioxide. As discussed later in this paper, silicone dioxide moderation does not represent a criticality threat. For graphite, both heterogeneous fuel arrangements and very large volume fractions of graphite are necessary for a graphite moderated system to go critical. Based on the observations and measurements of the FCM compositions, these conditions do not appear creditable for the Chernobyl FCM. Therefore, the focus of the analysis reported in this paper will be on reasonable heterogeneous fuel arrangements and water moderation. The analysis will evaluate a range of fuel and diluent compositions

Fuel element transport container with a removable cover

International Nuclear Information System (INIS)

Dannehl, G.; Fink, W.; Haenle, G.

1980-01-01

The cover of the fuel element transport container is removably fixed with screws on a flange as mechanical loads have to be expected during the transfer to the disposal plant. A ring-shaped or star-shaped clamping device grips over the cover. It has a clamp claw to lock the cover and permits unscrewing without unlocking the cover. (DG) [de

Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

International Nuclear Information System (INIS)

Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

2013-10-01

The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

Uncanistered Spent Nuclear fuel Disposal Container System Description Document

International Nuclear Information System (INIS)

Pettit, N. E.

2001-01-01

The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident

Quality assurance measures for spent fuel shipping and storage containers

International Nuclear Information System (INIS)

Droste, B.; Roedel, R.

1987-01-01

Quality assurance measures are to be applied in production and operation to ensure the approved fuel-element-container design specifications. The authors concentrate on the official regulations pertaining to the application of a quality assurance system, on the compliance with design specifications ensured by certified manufacturing tests and in-service inspections. For nodular-cast-iron container bodies, the authors demonstrate the procedure by presenting the contents of the materials data sheet characterizing the material, and the production and test sequence plan for container casting. In addition, they state the quality assurance requirements for interim-storage containers which transgress those stipulated for shipping containers. (orig.) [de

Procedure for filling with gas and sealing a nuclear fuel element consisting of a container

International Nuclear Information System (INIS)

Boyko, E.S.; Campbell, J.; Wiggins, R.J.

1971-01-01

A procedure for sealing the end plug of a fuel pin of a zirconium alloy or stainless steel within a pressure container, which contains an inert gas (preferably helium) atmosphere at a pressure of 35-133 kp/cm 2 , is described. The internal pressure in the fuel pin allows detection of leakages by means of a helium spectrometer and reduces the compressive stresses to which the fuel is subjected in the reactor. (JIW)

Nuclear fuel element containing particles of an alloyed Zr, Ti, and Ni getter material

International Nuclear Information System (INIS)

Grossman, L.N.; Levin, H.A.

1975-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles of alloy. The container is preferably held in the spring in the plenum of the fuel element. (Official Gazette)

49 CFR 173.230 - Fuel cell cartridges containing hazardous material.

Science.gov (United States)

2010-10-01

... integral vent feature) is deemed to pass the fire test if: (A) The internal pressure vents to zero gauge... required to power the equipment, plus two spares; (5) Large robust articles containing fuel cells may be...

Method for calculating the duration of vacuum drying of a metal-concrete container for spent nuclear fuel

Science.gov (United States)

Karyakin, Yu. E.; Nekhozhin, M. A.; Pletnev, A. A.

2013-07-01

A method for calculating the quantity of moisture in a metal-concrete container in the process of its charging with spent nuclear fuel is proposed. A computing method and results obtained by it for conservative estimation of the time of vacuum drying of a container charged with spent nuclear fuel by technologies with quantization and without quantization of the lower fuel element cluster are presented. It has been shown that the absence of quantization in loading spent fuel increases several times the time of vacuum drying of the metal-concrete container.

Application of fire-retardant treatment to the wood in Type A unirradiated nuclear fuel outer containers

International Nuclear Information System (INIS)

Whitlow, J.D.; Luna, R.E.

1992-01-01

Packagings for transporting unirradiated nuclear fuel assemblies in the United States are commonly constructed as rectangular boxes consisting of a metal inner container, a wooden outer container, and cushioning material separating the two. The wood in the outer container is a potential source of fuel for fire. Use of a fire-retardant treatment on the wood may reduce or eliminate the damage to nuclear fuel assemblies in some types of accidents involving fire. The applicability of using fire-retardant treatments on the wood of outer containers is addressed. An approximate cost-benefit analysis to determine if fire-retardant treatments are economically justified is presented. (Author)

Method of chemical reprocessing of irradiated nuclear fuels (especially fuels containing uranium)

International Nuclear Information System (INIS)

Koch, G.

1975-01-01

The invention deals with a method for the extraction especially of fast breeder fuels of high burn-up. A quaternary ammonium nitrate of high molecular weight is put into an organic diluting medium as extraction agent, corresponding to the general formula NRR'R''R'''NO 3 where R,R' and R'' are aliphatic radicals, R''' a methyl radical and the sum of the C atoms is greater than 16. After the extraction of the aqueous nitric acid containing nuclear fuel solution with this extracting agent, uranium, plutonium (or also thorium) can be found to a very high percentage in the organic phase and can be practically quantitatively back-extracted by means of diluted nitric acid, sulphuric acid or acetic acid. By using 30 volume percent tricapryl methyl ammonium nitrate in diethyl benzene for example, a distribution coefficient of 10.3 is obtained for uranium. (RB/LH) [de

A methodology to analyze the creep behaviour of nuclear fuel waste containers

Energy Technology Data Exchange (ETDEWEB)

Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1995-12-01

The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs.

A methodology to analyze the creep behaviour of nuclear fuel waste containers

International Nuclear Information System (INIS)

Dutton, R.

1995-12-01

The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs

Physical modeling of spent-nuclear-fuel container

Directory of Open Access Journals (Sweden)

Wang Liping

2012-11-01

Full Text Available A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container. In this physical simulation model, a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample, and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting. Also, a mould system was designed, in which changeable mould materials can be used for both the outside and inside moulds for different applications. The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained. Results show that for most isothermal planes, the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points, indicating that this new physical simulation model has high simulation accuracy, and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container, such as composition of ductile iron, the pouring temperature, the selection of mould material and design of cooling system. In addition, to maintain the spheroidalization of the ductile iron, the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.

Lift-based up-ender and methods using same to manipulate a shipping container containing unirradiated nuclear fuel

Science.gov (United States)

Nilles, Michael J.

2017-08-01

A shipping container containing an unirradiated nuclear fuel assembly is lifted off the ground by operating a crane to raise a lifting tool comprising a winch. The lifting tool is connected with the shipping container by a rigging line connecting with the shipping container at a lifting point located on the shipping container between the top and bottom of the shipping container, and by winch cabling connecting with the shipping container at the top of the shipping container. The shipping container is reoriented by operating the winch to adjust the length of the winch cabling so as to rotate the shipping container about the lifting point. Shortening the winch cabling rotates the shipping container about the lifting point from a horizontal orientation to a vertical orientation, while lengthening the winch cabling rotates the shipping container about the lifting point from the vertical orientation to the horizontal orientation.

Oxygen potential of a prototypic Mo-cermet fuel containing plutonium oxide

Energy Technology Data Exchange (ETDEWEB)

Miwa, Shuhei, E-mail: miwa.shuhei@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki, 311-1393 (Japan); Osaka, Masahiko [Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki, 311-1393 (Japan); Nozaki, Takahiro; Arima, Tatsumi; Idemitsu, Kazuya [Kyushu University, 744 Motooka Nishi-ku, Fukuoka, 819-0395 (Japan)

2015-10-15

Oxygen potential of a prototypic Mo-cermet fuel containing 50 vol.% PuO{sub 2−x} were investigated by the thermogravimetric analysis in the temperature range from 1273 K to 1473 K. It was shown that the oxygen potential and oxidation rate of the Mo-cermet were the same as those of pure PuO{sub 2−x} below the oxygen potential of Mo/MoO{sub 2} oxidation reaction. The same features of the Mo-cermet sample containing 50 vol.% PuO{sub 2−x} with those of pure PuO{sub 2−x} were discussed in terms of the microstructure. - Highlights: • Oxygen potential of Mo-cermet fuel was investigated by thermogravimetric analysis. • It was the same as that of pure PuO{sub 2−x} below the oxygen potential for Mo/MoO{sub 2}. • Gradual oxidation of Mo matrix occurred only above the oxygen potential for Mo/MoO{sub 2}. • Mo matrix and PuO{sub 2−x} in Mo-cermet fuel can thus be thermochemically individual.

Oxygen potential of a prototypic Mo-cermet fuel containing plutonium oxide

International Nuclear Information System (INIS)

Miwa, Shuhei; Osaka, Masahiko; Nozaki, Takahiro; Arima, Tatsumi; Idemitsu, Kazuya

2015-01-01

Oxygen potential of a prototypic Mo-cermet fuel containing 50 vol.% PuO_2_−_x were investigated by the thermogravimetric analysis in the temperature range from 1273 K to 1473 K. It was shown that the oxygen potential and oxidation rate of the Mo-cermet were the same as those of pure PuO_2_−_x below the oxygen potential of Mo/MoO_2 oxidation reaction. The same features of the Mo-cermet sample containing 50 vol.% PuO_2_−_x with those of pure PuO_2_−_x were discussed in terms of the microstructure. - Highlights: • Oxygen potential of Mo-cermet fuel was investigated by thermogravimetric analysis. • It was the same as that of pure PuO_2_−_x below the oxygen potential for Mo/MoO_2. • Gradual oxidation of Mo matrix occurred only above the oxygen potential for Mo/MoO_2. • Mo matrix and PuO_2_−_x in Mo-cermet fuel can thus be thermochemically individual.

46 CFR 182.460 - Ventilation of spaces containing machinery powered by, or fuel tanks for, gasoline.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Ventilation of spaces containing machinery powered by, or fuel tanks for, gasoline. 182.460 Section 182.460 Shipping COAST GUARD, DEPARTMENT OF HOMELAND..., gasoline. (a) A space containing machinery powered by, or fuel tanks for, gasoline must have a ventilation...

Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

International Nuclear Information System (INIS)

Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

1984-02-01

The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

Heat transfer coefficient for lead matrixing in disposal containers for used reactor fuel

International Nuclear Information System (INIS)

Mathew, P.M.; Taylor, M.; Krueger, P.A.

1985-02-01

In the Canadian Nuclear Fuel Waste Management Program, metal matrices with low melting points are being evaluated for their potential to provide support for the shell of disposal containers for used fuel, and to act as an additional barrier to the release of radionuclides. The metal matrix would be incorporated into the container by casting. To study the heat transfer processes during solidification, a steady-state technique was used, involving lead as the cast metal, to determine the overall heat transfer coefficient between the lead and some of the candidate container materials. The existence of an air gap between the cast lead and the container material appeared to control the overall heat transfer coefficient. The experimental observations indicated that the surface topography of the container material influences the heat transfer and that a smoother surface results in a greater heat transfer than a rough surface. The experimental results also showed an increasing heat transfer coefficient with increasing temperature difference across the container base plates; a model developed to base-plate bending can explain the observed results

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

Fuel salt and container material studies for MOSART transforming system

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

Fuel element transport container

International Nuclear Information System (INIS)

Benna, P.; Neuenfeldt, W.

1979-01-01

The reprocessing system includes a large number of waterfilled ponds next to each other for the intermediate storage of fuel elements from LWR's. The fuel element transport device is allocated to a middle pond. The individual ponds are separated from each other by walls, and are only accessible from the middle pond via narrow passages. The transport device includes a telescopic running rail for a trolley with a grab device for the fuel element. The running rail is supported in turn by a second trolley, which can be moved by wheels on rails. Part of the drive of the first trolley is arranged on the second one. Using this transport device, adjacent ponds can be served through the passage openings. (DG) [de

Design, fabrication and testing of a prototype stressed-shell fuel isolation container

International Nuclear Information System (INIS)

Crosthwaite, J.L.; Barrie, J.N.; Nuttall, K.

1982-07-01

Atomic Energy of Canada Limited is conducting and coordinating research into the development of engineered barriers for the disposal of unreprocessed irradiated fuel within a deep, stable geologic vault. In one approach, a containment shell of corrosion-resistant metal is proposed as the principal barrier to radionuclide release, giving a high probability of containment for at least 300 years, thus ensuring isolation of nearly all fission products for their hazardous lives. The simplest concept is the 'stressed-shell' container, designed with sufficient shell thickness to withstand the hydrostatic pressure within a 1000-m deep disposal vault postulated to have flooded with groundwater. This report describes the design, fabrication, analysis and hydrostatic testing of a full-scale stressed-shell prototype. The report concludes that the deformation and collapse performance of stressed-shell designs, based on short-term mechanical properties be modelled adequately by BOSOR 5, a commercially available stress-strain computer program. If the stressed-shell concept is retained as a viable fuel isolation concept, future analyses should include an assessment of the role of material creep on long-term container performance

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

Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

International Nuclear Information System (INIS)

Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

1996-12-01

In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

Plant for producing an oxygen-containing additive as an ecologically beneficial component for liquid motor fuels

Science.gov (United States)

Siryk, Yury Paul; Balytski, Ivan Peter; Korolyov, Volodymyr George; Klishyn, Olexiy Nick; Lnianiy, Vitaly Nick; Lyakh, Yury Alex; Rogulin, Victor Valery

2013-04-30

A plant for producing an oxygen-containing additive for liquid motor fuels comprises an anaerobic fermentation vessel, a gasholder, a system for removal of sulphuretted hydrogen, and a hotwell. The plant further comprises an aerobic fermentation vessel, a device for liquid substance pumping, a device for liquid aeration with an oxygen-containing gas, a removal system of solid mass residue after fermentation, a gas distribution device; a device for heavy gases utilization; a device for ammonia adsorption by water; a liquid-gas mixer; a cavity mixer, a system that serves superficial active and dispersant matters and a cooler; all of these being connected to each other by pipelines. The technical result being the implementation of a process for producing an oxygen containing additive, which after being added to liquid motor fuels, provides an ecologically beneficial component for motor fuels by ensuring the stability of composition fuel properties during long-term storage.

Evaluation of molten fuel containment concepts for gas-cooled fast breeder reactors

International Nuclear Information System (INIS)

Kang, C.S.; Torri, A.

1979-10-01

Four in-vessel molten fuel containment concepts for the GCFR were compared, namely, (1) a ceramic crucible, (2) a borax bath, (3) a heavy metal bath, and (4) a steel bath. The ceramic crucible is the simplest but depends on substantial upward heat removal. The borax bath and the heavy metal bath concepts offer better performance but would require design changes and an increased experimental effort. The steel bath concept is a good compromise and has potential for further improvement by combining it with the essential features of other concepts, i.e., the crucible or the heavy metal bath. It is concluded that several concepts could potentially exploit the normally provided cooled liner barrier in the PCRV cavity for post-accident fuel containment

Method for the chemical reprocessing of irradiated nuclear fuels, in particular nuclear fuels containing uranium

International Nuclear Information System (INIS)

Koch, G.

1976-01-01

In the chemical processing of irradiated uranium-containing nuclear fuels which are hydrolyzed with aqueous nitric acid, a suggestion is made to use as quaternary ammonium nitrate trialkyl-methyl ammonium nitrates as extracting agent, in which the sum of C atoms is greater than 16. In the illustrated examples, tricaprylmethylammonium nitrate, trilaurylmethylammonium nitrate and tridecylmethylammonium nitrate are named. (HPH/LH) [de

ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

International Nuclear Information System (INIS)

Wright, A.L.; Kress, T.S.; Smith, A.M.

1980-01-01

This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

Pre-conceptual design of a spent PWR fuel disposal container

International Nuclear Information System (INIS)

Choi, Jong Won; Cho, Dong Keun; Lee, Yang; Choi, Heui Joo; Lee, Jong Youl

2005-01-01

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid and bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert. the Maximum Von Mises stress from the 102 cm container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by ∼20 tons

Nitrogen oxides from combustion of nitrogen-containing polymers in waste-derived fuels

International Nuclear Information System (INIS)

Zevenhoven, R.; Kilpinen, P.; Hupa, M.; Elomaa, M.

2000-01-01

Usually, waste-derived fuels present nitrogen-containing fractions, which produce nitrogen oxides (NO) during combustion. This study was mainly concerned with poly amides (PA) (nylon), poly urethanes (PU), urea formaldehyde (UF) glue, sewage sludge and refuse-derived fuels (RDF). For control purposes, the authors chose a Polish sub-bituminous coal and a Finnish pine wood sample. An almost inverse trend between fuel nitrogen content and NO emissions was revealed through analysis of NO emissions at 850 Celsius, 1 bar, 7 per cent O 2 in N 2 . It was not possible to derive a clear correlation to the amount of ash generated by the samples. PU foam decomposed through a two-step process, as suggested by thermochromatography, and PA6-containing samples yielded epsilon-caprolactam as a major decomposition product. Important decomposition products from PU, PA6, PA6/PE, sewage sludge and UF glue samples were greenhouse gases as demonstrated by pyrolysis-gas chromatography/mass spectroscopy. The work was carried out at Abo Akademi University and University of Helsinki, Finland. 5 refs., 2 tabs., 3 figs

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

International Nuclear Information System (INIS)

Barranco R, F.

2015-01-01

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

EFFECT OF COMPOSITION OF FUEL CONTAINING BUTANOL ON WORKING PROCESS PARAMETERS OF DIESEL ENGINE

Directory of Open Access Journals (Sweden)

D. G. Hershan

2017-01-01

Full Text Available Computational researches the effect of composition of fuel containing butanol on working process parameters of 4ЧН 11/12,5 diesel engine on the external speed characteristic have been conducted. Nominal power is 140 kW at engine speed 2300 min–1. The engine is equipped with gas turbine pressure charging with intercooling of charging air, accumulator-type fuel-handling system. Calculations of the working process have been made in accordance with the developed computer program and models. Investigations have been carried out in two stages: without any changes in regulation of fuel-handling system and with cyclic fuel delivery that ensure such value of excess air factor at various operational modes which corresponds to the operation with diesel fuel. All the obtained results have been analyzed in the paper. The paper shows changes in mean indicated pressure, specific indicated fuel consumption, indicated efficiency, specific nitrogen oxides emissions for various modes in question while using 5, 10, 15, 20, 25 and 30 % mixture of diesel fuel with butanol. Dependences of parameters pertaining to diesel operation have been determined according to external speed characteristic for various mixtures and the obtained data make it possible to justify parameters of the fuel-handling system. It has been recommended to use a diesel fuel-butanol mixture containing 15 % of butanol without any changes in regulating and design engine parameters. It has been revealed that in order to improve parameters of the engine operational process mixture composition must be changed while changing the operational mode. An injector nozzle with a compound needle for the fuel-handling system has been developed and it allows to change fuel composition according to engine operational mode.

Assessment of materials for nuclear fuel immobilization containers

Energy Technology Data Exchange (ETDEWEB)

Nuttall, K; Urbanic, V F

1981-09-01

A wide range of engineering metals and alloys has been assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The expected range of service conditions in the disposal vault are discussed, as well as the material properties required for this application. An important requirement is that the container last at least 500 years without being breached. The assessment is treated in two parts. Part I concentrates on the physical and mechanical metallurgy, with special reference to strength, weldability, potential embrittlement mechanisms and some economic aspects. Part II discusses possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditions in the vault. Localized corrosion and delayed fracture processes are identified as being most likely to limit container lifetime. Hence an essential requirement is that such processes either be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further consideration as possible container materials: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the enviromental conditions in the vault. 42 figures, 31 tables.

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

Measurements on high temperature fuel cells with carbon monoxide-containing fuel gases; Messungen an Hochtemperatur-Brennstoffzellen mit kohlenmonoxidhaltigen Brenngasen

Energy Technology Data Exchange (ETDEWEB)

Apfel, Holger

2012-10-10

In the present work the different power density of anode-supported high-temperature solid oxide fuel cells (ASC-SOFCs) were examined for carbon monoxide-containing fuels. In addition to wet hydrogen / carbon monoxide mixtures the cells were run with synthetic gas mixtures resembling the products of an autothermal reformer, and actual reformate generated by a 2 kW autothermal reformer. It was found that the power-voltage characteristics of an ASC depends primarily on the open circuit voltages of different gas mixtures, but is nearly independent of the hydrogen concentration of the fuel, although the reaction rates of other potential fuels within the gas mixture, namely carbon monoxide and methane, are much lower that the hydrogen reaction rate. The probable reason is that the main fuel for the electrochemical oxidation within the cell is hydrogen, while the nickel in the base layer of the anode acts as a reformer which replenishes the hydrogen by water reduction via carbon monoxide and methane oxidation.

An assessment of materials for nuclear fuel immobilization containers

International Nuclear Information System (INIS)

Nuttall, K.; Urbanic, V.F.

1981-09-01

A wide range of engineering metals and alloys was assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The container must last at least 500 years without being breached. Materials were assessed for their physical and mechanical metallurgy, weldability, potential embrittlement mechanisms, and economics. A study of the possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditons in the vault showed that localized corrosion and delayed fracture processes are the most likely to limit container lifetime. Thus such processes either must be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further study: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the environmental conditions in the vault

Ceramic waste forms for fuel-containing masses at Chernobyl

International Nuclear Information System (INIS)

Oversby, V.M.

1994-05-01

The fuel materials originally in the core of the Chernobyl Unit 4 reactor are now present within the Ukrytie in three major forms: (1) very fine particles of fuel dispersed as dust (about 10 tonnes), (2) fragments of the destroyed core, and (3) lavas containing fuel, cladding, and other materials. All of these materials will need to be immobilized into waste forms suitable for final disposal. We propose a ceramic waste form system that could accommodate all three waste types with a single set of processing equipment. The waste form would include the mineral zirconolite for immobilization of actinide materials (including uranium), perovskite, nepheline, spinel, and other phases as dictated by the chemistry of the lava masses. Waste loadings as high as 50% U can be achieved if pyrochlore, a close relative of zirconolite, is used as the U host. The ceramic immobilization could be achieved with low additions of inert chemicals to minimize the final disposal volume while ensuring a durable product. The sequence of processing would be to collect and immobilize the fuel dust first. This material will require minimal preprocessing and will provide experience in the handling of the fuel materials. Core fragments would be processed next, using a cryogenic crushing stage to reduce the size prior to adding ceramic additives. The lavas would be processed last, which is compatible with the likely sequence of availability of materials and with the complexity of the operations. The lavas will require more adjustment of chemical additive composition than the other streams to ensure that the desired phases are produced in the waste form

The thermal analysis of BR-100: A barge/rail nuclear spent fuel transportation container

International Nuclear Information System (INIS)

Copsey, A.B.

1992-01-01

B ampersand W Fuel Company is designing a spent-fuel container called BR-100 that can be used for either barge or rail transport. This paper presents the thermal design and analysis. Both normal operation and hypothetical accident thermal transient conditions are evaluated. The BR-100 cask has a concrete layer than contains free water. During a hypothetical accident, the free water vaporizes and flows from the cask, removing a significant amount of thermal transient energy. The BR-100 transportation package meets the thermal requirements of 10CFR71. It additionally offers substantial margins to established material temperature limits

Regulated and unregulated emissions from an internal combustion engine operating on ethanol-containing fuels

Science.gov (United States)

Poulopoulos, S. G.; Samaras, D. P.; Philippopoulos, C. J.

In the present work, the effect of ethanol addition to gasoline on regulated and unregulated emissions is studied. A 4-cylinder OPEL 1.6 L internal combustion engine equipped with a hydraulic brake dynamometer was used in all the experiments. For exhaust emissions treatment a typical three-way catalyst was used. Among the various compounds detected in exhaust emissions, the following ones were monitored at engine and catalyst outlet: methane, hexane, ethylene, acetaldehyde, acetone, benzene, 1,3-butadiene, toluene, acetic acid and ethanol. Addition of ethanol in the fuel up to 10% w/w had as a result an increase in the Reid vapour pressure of the fuel, which indicates indirectly increased evaporative emissions, while carbon monoxide tailpipe emissions were decreased. For ethanol-containing fuels, acetaldehyde emissions were appreciably increased (up to 100%), especially for fuel containing 3% w/w ethanol. In contrast, aromatics emissions were decreased by ethanol addition to gasoline. Methane and ethanol were the most resistant compounds to oxidation while ethylene was the most degradable compound over the catalyst. Ethylene, methane and acetaldehyde were the main compounds present at engine exhaust while methane, acetaldehyde and ethanol were the main compounds in tailpipe emissions for ethanol fuels after the catalyst operation.

Developing an energy efficient steam reforming process to produce hydrogen from sulfur-containing fuels

Science.gov (United States)

Simson, Amanda

Hydrogen powered fuel cells have the potential to produce electricity with higher efficiency and lower emissions than conventional combustion technology. In order to realize the benefits of a hydrogen fuel cell an efficient method to produce hydrogen is needed. Currently, over 90% of hydrogen is produced from the steam reforming of natural gas. However, for many applications including fuel cell vehicles, the use of a liquid fuel rather than natural gas is desirable. This work investigates the feasibility of producing hydrogen efficiently by steam reforming E85 (85% ethanol/15% gasoline), a commercially available sulfur-containing transportation fuel. A Rh-Pt/SiO2-ZrO2 catalyst has demonstrated good activity for the E85 steam reforming reaction. An industrial steam reforming process is often run less efficiently, with more water and at higher temperatures, in order to prevent catalyst deactivation. Therefore, it is desirable to develop a process that can operate without catalyst deactivation at more energy efficient conditions. In this study, the steam reforming of a sulfur-containing fuel (E85) was studied at near stoichiometric steam/carbon ratios and at 650C, conditions at which catalyst deactivation is normally measured. At these conditions the catalyst was found to be stable steam reforming a sulfur-free E85. However, the addition of low concentrations of sulfur significantly deactivated the catalyst. The presence of sulfur in the fuel caused catalyst deactivation by promoting ethylene which generates surface carbon species (coke) that mask catalytic sites. The amount of coke increased during time on stream and became increasingly graphitic. However, the deactivation due to both sulfur adsorption and coke formation was reversible with air treatment at 650°C. However, regenerations were found to reduce the catalyst life. Air regenerations produce exotherms on the catalyst surface that cause structural changes to the catalyst. During regenerations the

Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

International Nuclear Information System (INIS)

Matos, J.E.

1994-12-01

This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy's Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006

Cold spray copper coatings for used fuel containers

Energy Technology Data Exchange (ETDEWEB)

Keech, P. [Nuclear Waste Management Organization, Toronto, ON (Canada); Vo, P.; Poirier, D.; Legoux, J-G [National Research Council, Boucherville QC, (Canada)

2015-07-01

Recently, the Nuclear Waste Management Organization has been developing copper coatings as a method of protecting steel used fuel containers (UFCs) from corrosion within a deep geological repository. The corrosion barrier design is based on the application of a copper coating bonded directly to the exterior surface of the UFC structural core. Copper coating technologies amendable to supply of pre-coated UFC vessel components and application to the weld zone following UFC closure within the radiological environment have been investigated. Copper cold spray has been assessed for both operations; this paper outlines the research and development to date of this technique. (author)

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

International Nuclear Information System (INIS)

Travelli, A.

1988-01-01

A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface

Method for forming nuclear fuel containers of a composite construction and the product thereof

International Nuclear Information System (INIS)

Cheng, B.-C.; Rosenbaum, H.S.; Armijo, J.S.

1981-01-01

An improved method of producing a composite nuclear fuel container is described which comprises a casing or fuel sheath of zirconium or its alloy with a lining cladding of deposited copper superimposed over the inside surface of the zirconium or alloy and a layer of oxide of the zirconium or alloy formed on the inside surface of the casing or sheath. (U.K.)

Progress in welding studies for Canadian nuclear fuel waste disposal containers

International Nuclear Information System (INIS)

Maak, P.Y.Y.

1985-11-01

This report describes the progress in the development of closure-welding technology for Canadian nuclear fuel waste disposal containers. Titanium, copper and Inconel 625 are being investigated as candidate materials for fabrication of these containers. Gas-tungsten-arc welding, gas metal-arc-welding, resistance-heated diffusion bonding and electron beam welding have been evaluated as candidate closure welding processes. Characteristic weldment properties, relative merits of welding techniques, suitable weld joint configurations and fit-up tolerances, and welding parameter control ranges have been identified for various container designs. Furthermore, the automation requirements for candidate welding processes have been assessed. Progress in the development of a computer-controlled remote gas-shielded arc welding system is described

The Fuel Performance Analysis of LWR Fuel containing High Thermal Conductivity Reinforcements

International Nuclear Information System (INIS)

Kim, Seung Su; Ryu, Ho Jin

2015-01-01

The thermal conductivity of fuel affects many performance parameters including the fuel centerline temperature, fission gas release and internal pressure. In addition, enhanced safety margin of fuel might be expected when the thermal conductivity of fuel is improved by the addition of high thermal conductivity reinforcements. Therefore, the effects of thermal conductivity enhancement on the fuel performance of reinforced UO2 fuel with high thermal conductivity compounds should be analyzed. In this study, we analyzed the fuel performance of modified UO2 fuel with high thermal conductivity reinforcements by using the FRAPCON-3.5 code. The fissile density and mechanical properties of the modified fuel are considered the same with the standard UO2 fuel. The fuel performance of modified UO2 with high thermal conductivity reinforcements were analyzed by using the FRAPCON-3.5 code. The thermal conductivity enhancement factors of the modified fuels were obtained from the Maxwell model considering the volume fraction of reinforcements

Process and container system for transferring or transporting fuel elements from a nuclear power station to a store

International Nuclear Information System (INIS)

Vox, A.J.

1984-01-01

A system of containers with three types of containers (an inside container, a transport container and a storage container) is used. One either sets the inside container open on the lid side into the transport container first in the water pond of the nuclear power station, and one then sets the fuel elements into the inside container, or one places the inside container, loaded with fuel elements away from the transport container, into the transport container. Both containers are then closed and are transported to the store as a unit. The storage container open on the lid side is prepared there, the floor of the transport container is opened and this, together with the inside container, is lifted above the storage container or set above the storage container. The inside container is then lowered onto the storage container, the transport container is removed and the lid of the storage container is closed. (orig./HP) [de

10 CFR 503.38 - Permanent exemption for certain fuel mixtures containing natural gas or petroleum.

Science.gov (United States)

2010-01-01

... natural gas or petroleum. 503.38 Section 503.38 Energy DEPARTMENT OF ENERGY (CONTINUED) ALTERNATE FUELS... mixtures containing natural gas or petroleum. (a) Eligibility. Section 212(d) of the Act provides for a... proposes to use a mixture of natural gas or petroleum and an alternate fuel as a primary energy source; (2...

Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations

International Nuclear Information System (INIS)

Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J.

1991-12-01

This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 x 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990

The single SNR fuel assembly container (ESBB) to transport unirradiated SNR 300 fuel assemblies

International Nuclear Information System (INIS)

Hilbert, F.; Hottenrott, G.

1998-01-01

In this paper a new type B(U) package design is presented. The Single SNR Fuel Assembly Container (ESBB) is designed for the transport and storage of a single SNR 300 fuel assembly. This package is the main component for the future interim storage of the fuel assemblies in heavy storage casks. Its benefits are that it is compatible with the Category I transport system of Nuclear Cargo + Service NCS) used in Germany and that it can be easily handled at the current storage locations as well as in an interim storage facility. In total 205 fuel assemblies are currently stored in Hanau, Germany and Dounreay, U.K. Former studies have shown, that heavy transport and storage casks can be handled there only with considerable efforts. But the required category I transport to an interim storage is not reasonably feasible. To overcome these problems the ESBB was designed. It consists of a stainless steel tube with welded bottom, a welded plug as closure system and shock absorbers 26 packages at maximum can be transported in one batch with the NCS security vehicle. The safety analysis shows that the package complies with IAEA 1996. Standard calculations methods and computer codes like HEATING 7.2 (Childs 1993) have been used for the analysis. Criticality safety assessment is based on conservative assumptions as required in IAEA 1996. Drop tests carried out by BAM will be used to verify the design. These tests are scheduled for mid 1998. For the validation of the design prototypes have already been manufactured. Handling tests show that the design complies with the requirements. Preliminary drop tests show that the certification drop tests will be passed positively. (authors)

Analysis to the criticality the storage and containers to the Juragua Nuclear Power Plant Fuel

International Nuclear Information System (INIS)

Guerra Valdes, R.

1998-01-01

Presently analysis the criticality the warehouses and containers the nuclear fuels in Juragua nuclear power plant the property multiplicity determined in these system and it is verified that for the geometry and operation conditions defined in the design as well as in accidents situations, the arrangement the fuel stays subcritical with an appropriate margin

Method of burning sulfur-containing fuels in a fluidized bed boiler

Science.gov (United States)

Jones, Brian C.

1982-01-01

A method of burning a sulfur-containing fuel in a fluidized bed of sulfur oxide sorbent wherein the overall utilization of sulfur oxide sorbent is increased by comminuting the bed drain solids to a smaller average particle size, preferably on the order of 50 microns, and reinjecting the comminuted bed drain solids into the bed. In comminuting the bed drain solids, particles of spent sulfur sorbent contained therein are fractured thereby exposing unreacted sorbent surface. Upon reinjecting the comminuted bed drain solids into the bed, the newly-exposed unreacted sorbent surface is available for sulfur oxide sorption, thereby increasing overall sorbent utilization.

Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

International Nuclear Information System (INIS)

Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

2007-01-01

Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

Energy Technology Data Exchange (ETDEWEB)

De Wild, P.J.; Nyqvist, R.G.; De Bruijn, F.A.; Stobbe, E.R. [ECN Hydrogen and Clean Fossil Fuels, Petten (Netherlands)

2006-02-15

Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC)) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20 ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG.

Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

Energy Technology Data Exchange (ETDEWEB)

de Wild, P.J.; Nyqvist, R.G.; de Bruijn, F.A.; Stobbe, E.R. [Energy Research Centre of The Netherlands ECN, P.O. Box 1, 1755 ZG Petten (Netherlands)

2006-09-22

Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG. (author)

CLASSIFICATION OF THE MGR NON-FUEL COMPONENTS DISPOSAL CONTAINER SYSTEM

International Nuclear Information System (INIS)

J.A. Ziegler

1999-01-01

The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) non-fuel components disposal container system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

Non-traditional Process of Hydrogen Containing Fuel Mixtures Production for Internal-combustion Engines

Directory of Open Access Journals (Sweden)

Gennady G. Kuvshinov

2012-12-01

Full Text Available The article justifies the perspectives of development of the environmentally sound technology of hydrogen containing fuel mixtures for internal-combustion engines based on the catalytic process of low-temperature decomposition of hydrocarbons into hydrogen and nanofibrous carbon.

Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

International Nuclear Information System (INIS)

Johnson, G.L.

1988-09-01

Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97/degree/C and whether the cladding of the stored spent fuel ever exceeds 350/degree/C. Limiting the borehole to temperatures of 97/degree/C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350/degree/C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97/degree/C for the full 1000-yr analysis period

Department of the Navy final environmental impact statement for a container system for the management of naval spent nuclear fuel

International Nuclear Information System (INIS)

1996-11-01

This Final Environmental Impact Statement (EIS) addresses six general alternative systems for the loading, storage, transport, and possible disposal of naval spent nuclear fuel following examination. This EIS describes environmental impacts of (1) producing and implementing the container systems (including those impacts resulting from the addition of the capability to load the containers covered in this EIS in dry fuel handling facilities at Idaho National Engineering Laboratory (INEL)); (2) loading of naval spent nuclear fuel at the Expended Core Facility or at the Idaho Chemical Processing Plant with subsequent storage at INEL; (3) construction of a storage facility (such as a paved area) at alternative locations at INEL; and (4) loading of containers and their shipment to a geologic repository or to a centralized interim storage site outside the State of Idaho once one becomes available. As indicated in the EIS, the systems and facilities might also be used for handling low-level radiological waste categorized as special case waste. The Navy's preferred alternative for a container system for the management of naval spent fuel is a dual-purpose canister system. The primary benefits of a dual-purpose canister system are efficiencies in container manufacturing and fuel reloading operations, and potential reductions in radiation exposure

Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

Science.gov (United States)

Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

Fuel containment and damage tolerance for large composite primary aircraft structures. Phase 1: Testing

Science.gov (United States)

Sandifer, J. P.

1983-01-01

Technical problems associated with fuel containment and damage tolerance of composite material wings for transport aircraft were identified. The major tasks are the following: (1) the preliminary design of damage tolerant wing surface using composite materials; (2) the evaluation of fuel sealing and lightning protection methods for a composite material wing; and (3) an experimental investigation of the damage tolerant characteristics of toughened resin graphite/epoxy materials. The test results, the test techniques, and the test data are presented.

Structural design of a shipping container of fuel elements, non-irradiated, for research reactors

International Nuclear Information System (INIS)

Morales Uzqueda, Eduardo Mario

2013-01-01

This work is part of a project whose ultimate goal the creation and subsequent discharge of a transport container fuel assemblies for use by the Chilean Commission for Energy Nuclear. In principle it is covered in the design stage, considering the materials and methods used, to further develop a stage of checking voltages in the container to be manufactured. To achieve the first phase of the study is necessary to understand and warn the importance, geometry and content of the fuel elements to be transported, for which there are standards that provide fundamental material for proper classification of both content and container design. Once approved the design of the structure is critical examine both in normal operation and in the case of accidents that are established by international bodies. for appropriate analytical methods that seek to achieve is use a appropriate representation of the behavior of the structure. in addition to strengthen the theory computer simulations of the tests used applied, where the results will be contrasted with the first method of calculation. Results are obtained for the stress field and displacement total delivering the information necessary to approve the container

Calculation of the process of vacuum drying of a metal-concrete container with spent nuclear fuel

Science.gov (United States)

Karyakin, Yu. E.; Lavrent'ev, S. A.; Pavlyukevich, N. V.; Pletnev, A. A.; Fedorovich, E. D.

2012-01-01

An algorithm and results of calculation of the process of vacuum drying of a metal-concrete container intended for long-term "dry" storage of spent nuclear fuel are presented. A calculated substantiation of the initial amount of moisture in the container is given.

Effect of buoyancy on fuel containment in an open-cycle gas-core nuclear rocket engine.

Science.gov (United States)

Putre, H. A.

1971-01-01

Analysis aimed at determining the scaling laws for the buoyancy effect on fuel containment in an open-cycle gas-core nuclear rocket engine, so conducted that experimental conditions can be related to engine conditions. The fuel volume fraction in a short coaxial flow cavity is calculated with a programmed numerical solution of the steady Navier-Stokes equations for isothermal, variable density fluid mixing. A dimensionless parameter B, called the Buoyancy number, was found to correlate the fuel volume fraction for large accelerations and various density ratios. This parameter has the value B = 0 for zero acceleration, and B = 350 for typical engine conditions.

Analysis of the ecological parameters of the diesel engine powered with biodiesel fuel containing methyl esters from Camelina sativa oil

Directory of Open Access Journals (Sweden)

S. Lebedevas

2010-03-01

Full Text Available The article explores the possibilities of using fatty acid methyl esters derived from the oil of a new species of oily plant Camelina sativa not demanding on soil. The performed research on the physical and chemical properties of pure methyl esters from Camelina sativa show that biofuels do not meet requirements for the biodiesel fuel standard (LST EN 14214:2009 of a high iodine value and high content of linoleic acid methyl ester, so they must be mixed with methyl esters produced from pork lard the content of which in the mixture must be not less than 32%. This article presents the results of tests on combustion emission obtained when three-cylinder diesel engine VALMET 320 DMG was fuelled with a mixture containing 30% of this new kind of fuel with fossil diesel fuel comparing with emissions obtained when the engine was fuelled with a fuel mixture containing 30% of conventional biodiesel fuel (rapeseed oil methyl esters with fossil diesel fuel. The obtained results show that using both types of fuel, no significant differences in CO and NOx concentrations were observed throughout the tested load range. When operating on fuels containing methyl esters from Camelina sativa, HC emissions decreased by 10 to 12% and the smokeness of exhaust gas by 12 to 25%.

Management and Handling of Rejected Fuel of MTR Type and Process Effluents Contained Uranium at FEPI

International Nuclear Information System (INIS)

Ghaib Widodo; Bambang Herutomo

2007-01-01

Research Reactor Fuel Element Production Installation (FEPI) - Serpong has performed management and handling of all kinds of rejected fuel material during production (solids, liquids, and gases) and process effluents contained uranium. The methods that has been implemented are precipitation, absorption, evaporation, electrolysis, and electrodialysis. By these methods will finally be obtained forms of product which can be used directly as fuel material feed and solid/liquid radioactive waste that fulfil the requirements (uranium contents < 50 ppm) to be send to Radioactive Waste Management Installation. (author)

On material and energy sources of formation of fuel-containing materials during Chernobyl NPP UNIT 4 accident

Directory of Open Access Journals (Sweden)

O. V. Mikhailov

2016-12-01

Full Text Available Results of detailed analysis of material substance of lava-like fuel-containing materials sources (FCM and clusters with high uranium concentration were presented. Material and energy balance are aggregated in a process model for optimal composition of sacrificial materials and FCM. Quantitative estimate is given for spent nuclear fuel’ afterheat in a number of other heat energy sources in reactor vault. Conclusion was made that upon condition of 50 % heat loss, remained amount of “useful” heat would be sufficient for proceeding of blast furnace version of fuel-containing materials.

Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

International Nuclear Information System (INIS)

Johnson, G.L.

1991-11-01

Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

Fuels Containing Methane of Natural Gas in Solution

Science.gov (United States)

Sullivan, Thomas A.

2004-01-01

While exploring ways of producing better fuels for propulsion of a spacecraft on the Mars sample return mission, a researcher at Johnson Space Center (JSC) devised a way of blending fuel by combining methane or natural gas with a second fuel to produce a fuel that can be maintained in liquid form at ambient temperature and under moderate pressure. The use of such a blended fuel would be a departure for both spacecraft engines and terrestrial internal combustion engines. For spacecraft, it would enable reduction of weights on long flights. For the automotive industry on Earth, such a fuel could be easily distributed and could be a less expensive, more efficient, and cleaner-burning alternative to conventional fossil fuels. The concept of blending fuels is not new: for example, the production of gasoline includes the addition of liquid octane enhancers. For the future, it has been commonly suggested to substitute methane or compressed natural gas for octane-enhanced gasoline as a fuel for internal-combustion engines. Unfortunately, methane or natural gas must be stored either as a compressed gas (if kept at ambient temperature) or as a cryogenic liquid. The ranges of automobiles would be reduced from their present values because of limitations on the capacities for storage of these fuels. Moreover, technical challenges are posed by the need to develop equipment to handle these fuels and, especially, to fill tanks acceptably rapidly. The JSC alternative to provide a blended fuel that can be maintained in liquid form at moderate pressure at ambient temperature has not been previously tried. A blended automotive fuel according to this approach would be made by dissolving natural gas in gasoline. The autogenous pressure of this fuel would eliminate the need for a vehicle fuel pump, but a pressure and/or flow regulator would be needed to moderate the effects of temperature and to respond to changing engine power demands. Because the fuel would flash as it entered engine

Modeling of combustion products composition of hydrogen-containing fuels

International Nuclear Information System (INIS)

Assad, M.S.

2010-01-01

Due to the usage of entropy maximum principal the algorithm and the program of chemical equilibrium calculation concerning hydrogen--containing fuels are devised. The program enables to estimate the composition of combustion products generated in the conditions similar to combustion conditions in heat engines. The program also enables to reveal the way hydrogen fraction in the conditional composition of the hydrocarbon-hydrogen-air mixture influences the harmful components content. It is proven that molecular hydrogen in the mixture is conductive to the decrease of CO, CO 2 and CH x concentration. NO outlet increases due to higher combustion temperature and N, O, OH concentrations in burnt gases. (authors)

Fuel containing vessel for transporting nuclear fuel

International Nuclear Information System (INIS)

Yoshizawa, Hiroyasu; Shimizu, Fukuzo; Tanaka, Nobuyuki.

1996-01-01

A shock absorbing mechanism is disposed on an inner bottom of a vessel main body. The shock absorbing mechanism comprises a shock absorbing member disposed on the upper surface of a bottom wall, an annular metal plate disposed on the upper surface of the shock absorbing member and an annular spacer disposed on the upper surface of the metal plate. The shock absorbing member is made of a material such as of wood, lead, metal honeycomb or a metal mesh, which plastically deforms when applied with load higher than a predetermined level, and is formed in a square block-like form covering the upper surface of the bottom wall. The spacer is made of a thin soft material such as tetrafluoroethylene, and is formed in such a shape as capable of preventing direct contact of the lower end of the cylindrical member in a lower tie plate of nuclear fuels with the metal portion. This can ensure integrity of nuclear fuels even when they fall from a high place upon an assumed dropping accident. (I.N.)

Foreign research reactor spent nuclear fuel inventories containing HEU and LEU of US-origin

International Nuclear Information System (INIS)

Matos, J.E.

1995-01-01

This paper provides estimates of the quantities and types of foreign research reactor spent nuclear fuel containing HEU and LEU of US-origin that are anticipated during the period beginning in January 1996 and extending for 10-15 years

Considerations in modelling the melting of fuel containing fission products and solute oxides

International Nuclear Information System (INIS)

Akbari, F.; Welland, M.J.; Lewis, B.J.; Thompson, W.T.

2005-01-01

It is well known that the oxidation of a defected fuel element by steam gives rise to an increase in O/U ratio with a consequent lowering of the incipient melting temperature. Concurrently, the hyperstoichiometry reduces the thermal conductivity thereby raising the centerline fuel pellet temperature for a fixed linear power. The development of fission products soluble in the UO 2 phase or, more important, the deliberate introduction of additive oxides in advanced CANDU fuel bundle designs further affects and generally lowers the incipient melting temperature. For these reasons, the modeling of the molten (hyperstoichiometric) UO 2 phase containing several solute oxides (ZrO 2 , Ln 2 O 3 and AnO 2 ) is advancing in the expectation of developing a moving boundary heat and mass transfer model aimed at better defining the limits of safe operating practice as burnup advances. The paper describes how the molten phase stability model is constructed. The redistribution of components across the solid-liquid interface that attends the onset of melting of a non-stoichiometric UO 2 containing several solutes will be discussed. The issues of how to introduce boundary conditions into heat transfer calculations consistent with the requirements of the Phase Rule will be addressed. The Stefan problem of a moving boundary associated with the solid/liquid interface sets this treatment apart from conventional heat and mass transfer problems. (author)

Problems of interaction between water and fuel containing masses inside the object 'Shelter' of Chernobyl nuclear power plant

International Nuclear Information System (INIS)

Yukhnovs'kij, Yi.R.; Kobrin, O.Je.; Tokarchuk, M.V.; Tokarevs'kij, V.V.

1997-01-01

The main forms of the existence of nuclear fuel and major concomitant factors of nuclear and ecological danger of the object 'Shelter' are presented. The processes of interaction between water and fuel containing materials have been analysed on the basis of experimental data

Safety Analysis Report for Packaging: The unirradiated fuel shipping container USA/9853/AF

International Nuclear Information System (INIS)

1991-01-01

The HFBR Unirradiated Fuel Shipping Container was designed and fabricated at the Oak Ridge National Laboratory in 1978 for the transport of fuel for the High Flux Beam Reactor (HFBR) for Brookhaven National Laboratory. The package has been evaluated analytically, as well as the comparison to tests on similar packages, to demonstrate compliance with the applicable regulations governing packages in which radioactive and fissile materials are transported. The contents of this Safety Analysis Report for Packaging (SARP) are based on Regulatory Guide 7.9 (proposed Revision 2 - May 1986), 10 CFR Part 71, DOE Order 1540.2, DOE Order 5480.3, and 49 CFR Part 173

Optimal Protection of Reactor Hall Under Nuclear Fuel Container Drop Using Simulation Methods

Directory of Open Access Journals (Sweden)

Králik Juraj

2014-12-01

Full Text Available This paper presents of the optimal design of the damping devices cover of reactor hall under impact of nuclear fuel container drop of type TK C30. The finite element idealization of nuclear power plant structure is used in software ANSYS. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall in comparison with the experimental results. The probabilistic and sensitivity analysis of the damping devices was considered on the base of the simulation methods in program AntHill using the Monte Carlo method.

Proceedings of a workshop on corrosion of Nuclear fuel waste containers

International Nuclear Information System (INIS)

Shoesmith, D.W.

1990-01-01

The 23 papers presented at this conference review the technical merits, and particularly corrosion performance, of the three main materials used for nuclear fuel waste containers: titanium and its alloys, copper and its alloys, and iron and carbon steels. The specific questions posed to the Workshop were: 1) Can we predict the lifetime of container materials in a variety of vault environments? 2) Is there a limiting range of conditions beyond which a specific material cannot be used? 3) Do we have the necessary corrosion rate data and/or mechanistic models required to make predictions? 4) Can we justify the use of titanium on the basis of propagation rate measurements for crevice corrosion, or do we need to prove initiation cannot occur? 5) Will the pitting of copper be significant? 6) How thick a carbon steel container would be required, and can it be fabricated and stress-relieved? 7) Are radiation fields of any consequence at the dose rates expected?

Group constants calculation for fuel assemblies containing burnable absorbers; Prorachun grupnih konstanti gorivnih elemenata koji sadrzhe sagorive apsorbere

Energy Technology Data Exchange (ETDEWEB)

Petrovic, B [Institut Rudjer Boskovic, Zagreb (Yugoslavia); Pevec, D [Elektrotehnicki Fakultet, Zagreb Univ. (Yugoslavia); Urli, N; Shmuc, T [Institut Rudjer Boskovic, Zagreb (Yugoslavia)

1988-07-01

The upgrading of the computer code package PSU-LEOPARD/MCRAC is described. The upgraded package enables modelling of fuel assemblies containing burnable absorbers in the form of borosilicate glass rodlets, or, integral fuel burnable absorbers. The package is tested using the NPP Krsko core data. (author)

Conditioning of spent fuel assemblies from the Rossendorf RFR research reactor in transport and storage containers of the type CASTOR MTR 2

International Nuclear Information System (INIS)

Schneider, B.; Hofmann, G.

1994-09-01

Most of the spent fuel assemblies are temporarily stored in the flooded fuel ponds AB 1 and AB 2 of the RFR, and some are still in the reactor core. The conditioning task described here is part of the RFR spent fuel management concept and covers the safe emplacement of the spent fuel elements in the CASTOR MTR 2 shipping containers and the sealing of the containers in compliance with the nuclear licence issued for the conditioning task. The transfer of the spent fuel assemblies from the present wet storage conditions to the dry storage conditions in the CASTOR MTR 2 containers is done by a mobile manipulation equipment consisting essentially of the transfer sluice gate and a transfer container. Subsequent to conditioning, the shipping containers are to be transported to a licensed intermediate storage facility to await their transport to a national radwaste repository. The technical handling tools for the transfer and manipulation are briefly described, as well as the process steps involved, putting emphasis on the detailed description of processes and the accompanying time frame, so that the conditioning task can be incorporated into the work plan of the entire project. The report further presents the EDP concept established for the task, including the required data archivation and documentation. (orig.) [de

Experimental verification of methods for gamma dose rate calculations in the vicinity of containers with the RA reactor spent fuel elements

International Nuclear Information System (INIS)

Milosevic, M.; Cupac, S.; Pesic, M.

2005-01-01

The methodology for equivalent gamma dose rate determination on the outer surface of existing containers with the spent fuel elements of the RA reactor is briefly summarised, and experimental verification of this methodology in the field of gamma rays near the aluminium channel with spent fuel elements lifted from the stainless steel containers no. 275 in the RA reactor hall is presented. The proposed methodology is founded on: the existing fuel burnup data base; methods and models for the photon source determination in the RA reactor spent fuel elements developed in the Vinca Institute, and validated Monte Carlo codes for the equivalent gamma dose rate calculations. (author) [sr

Evaluation of the in pile performance of boron containing fuel pellets

International Nuclear Information System (INIS)

Jeong, Gwanyoon; Sohn, Dongseong

2012-01-01

The world rare earth resource are heavily concentrated in certain area and if these natural resources are weaponized by a country, we may confront serious difficulty because rare earth element gadolinium(Gd) is used as burnable poison material in some nuclear power plants (NPP) in Korea. Gd is used as a neutron absorbing material in Gd 2 O 3 form and mixed with UO 2 When boron is used as burnable poison in nuclear fuel, in fuel pellets. The burnable poison mixed in the fuel pellets is called integral burnable absorber (BA) design which differentiates it from the old separate BA design. In the old separate BA design, boron(B) was used in borosilicate glass (PYREX) form and placed in guide tubes. With the development of the concern over the availability of rare earth material Gd, B is considered as a candidate material replacing Gd for the case when the rare earth material is weaponized. However the idea for new boron BA design is integral type because the integral type BA design has several benefits over the separate BA design, such as reduction of radioactive waste, more positions for BA location, etc. 10 B absorbs a neutron and produces helium by the following reaction: 10 B + n → 7 Li + 4 He The helium produced by the nuclear reaction may cause the increase of rod internal pressure and change the gap conductivity if the significant amount of helium gas is released to the gap between the pellet and the cladding. Thus, it is necessary to investigate the in-pile behaviors of B containing pellet. However, few experiment have been carried out so far on the behavior of in-pile produced helium in UO 2 fuel pellets, especially for the cases boron compound is mixed with UO 2 In this paper, we will evaluate the production and the release of helium depending on fuel. 10 B concentration in the fuel

Long time storage containers for spent fuels and vitrified wastes: synthesis of the studies

International Nuclear Information System (INIS)

Beziat, A.

2004-01-01

This report presents a synthesis of the studies relatives to the containers devoted to the long time spent fuels storage and vitrified wastes packages. These studies were realized in the framework of the axis 3 of the law of 1991 on the radioactive wastes management. The first part is devoted to the presentation of the studies. The container sizing studies which constitute the first containment barrier are then presented. The material choice and the closed system are also detailed. The studies were validate by the realization of containers models and an associated demonstration program is proposed. A synthesis of the technical and economical studies allowed to determine the components and operation costs. (A.L.B.)

Combined cogeneration equipment containing gas turbine using low sulphur heavy stock as fuel

Energy Technology Data Exchange (ETDEWEB)

Taguchi, Goro; Ishiki, Katsuhiko

1988-03-10

This paper describes the combined cogeneration in Chemical and Plastics Co. Madras (India) which uses low sulphur heavy stock (LSHS) as a fuel. By the combined cogeneration of gas turbine and boiler steam turbine power generation, the exhaust from the steam turbine is supplied to the factory as a process steam. This equipment has a capacity of 4835 kW in overall generation power and 23.5 tons/hrs. in steam evaporation. The gas turbine system is equipped with an axial-flow, 11 step compressor, an axial flow, 4 step turbine, and a single-can back flow combustor fixed to the intermediate casing. The temperature of the exhaust from the gas turbine is 542/sup 0/C. Low quality LSHS when burned exerts no influence on the service life of the turbine blades. The boiler is a horizontal bent pipe, forced circulation type, and the steam turbine is a back pressure control type. The fuel is treated with a horizontal, two drum, electrostatic separator to which a demulsifier is supplied, to be separated into oil and water. As to the vanadium salts contained in the fuels, a chemical liquid containing MgO as a major ingredient is added to the fuel prior to the combustion. Thereby, the melting temperature of the vanadium oxide is enhanced, which serves for prevention of the melting and adhesion of the vanadium oxide to the gas turbine. LSHS is a residual oil produced by the ordinary pressure distillation of India-produced crude oil, has a sulphur content of 1.75%, and is solid at room temperature. Attention should be paid to clogging of the pipings. The overall efficiency is 80%. The combined cogeneration can be coordinated with load variations of 10 - 20%. (12 figs, 1 tab)

Production of JET fuel containing molecules of high hydrogen content

Directory of Open Access Journals (Sweden)

Tomasek Sz.

2017-12-01

Full Text Available The harmful effects of aviation can only be reduced by using alternative fuels with excellent burning properties and a high hydrogen content in the constituent molecules. Due to increasing plastic consumption the amount of the plastic waste is also higher. Despite the fact that landfill plastic waste has been steadily reduced, the present scenario is not satisfactory. Therefore, the aim of this study is to produce JET fuel containing an alternative component made from straight-run kerosene and the waste polyethylene cracking fraction. We carried out our experiments on a commercial NiMo/Al2O3/P catalyst at the following process parameters: T=200-300°C, P=40 bar, LHSV=1.0-3.0 h-1, hydrogen/hydrocarbon ratio= 400 Nm3/m3. We investigated the effects of the feedstocks and the process parameters on the product yields, the hydrodesulfurization and hydrodearomatization efficiencies, and the main product properties. The liquid product yields varied between 99.7-99.8%. As a result of the hydrogenation the sulfur (1-1780 mg/kg and the aromatic contents (9.0-20.5% of the obtained products and the values of their smoke points (26.0-34.7 mm fulfilled the requirements of JET fuel standard. Additionally, the concentration of paraffins increased in the products and the burning properties were also improved. The freezing points of the products were higher than -47°C, therefore product blending is needed.

BE (fuel element)/ZL (interim storage facility) module. Constituents of the fuel BE data base for BE documentation with respect to the disposal planning and the support of the BE container storage administration

International Nuclear Information System (INIS)

Hoffmann, V.; Deutsch, S.; Busch, V.; Braun, A.

2012-01-01

The securing of spent fuel element disposal from German nuclear power plants is the main task of GNS. This includes the container supply and the disposal analysis and planning. Therefore GNS operates a data base comprising all in Germany implemented fuel elements and all fuel element containers in interim storage facilities. With specific program modules the data base serves an optimized repository planning for all spent fuel elements from German NPPS and the supply of required data for future final disposal. The data base has two functional models: the BE (fuel element) and the ZL (interim storage) module. The contribution presents the data structure of the modules and details of the data base operation.

Used fuel packing plant for CANDU fuel

Energy Technology Data Exchange (ETDEWEB)

Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

2015-07-01

Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

An assessment of the feasibility of indefinite containment of Canadian nuclear fuel wastes

International Nuclear Information System (INIS)

Shoesmith, D.W.; King, F.; Ikeda, B.M.

1995-05-01

This report presents an analysis of the expected corrosion behaviour of nuclear fuel waste containers in a conceptual Canadian disposal vault. The container materials considered are dilute Ti alloys (Grades-2, -12 and -16) and oxygen-free copper. The corrosive conditions within the disposal vault change with time as the initially trapped oxygen is consumed and as the heat and γ-radiation produced by the waste decays. This evolution of the vault environment is broadly classified into an early, warm and oxidizing period followed by a period of long-term, stable, cool and non-oxidizing conditions. The corrosion behaviour of both types of material during these two periods is discussed, and various models that have been developed to predict the lifetimes of the containers are presented. The conclusion is that indefinite containment of the waste is feasible with both copper and titanium alloys under Canadian disposal conditions. (author). refs., tabs., figs

Fuel composition effect on the electrostatically-driven atomization of bio-butanol containing engine fuel blends

International Nuclear Information System (INIS)

Agathou, Maria S.; Kyritsis, Dimitrios C.

2012-01-01

Highlights: ► Sprays of alcohol-containing blends are amenable to electrostatic manipulation. ► Monodispersion is non-achievable for conditions pertaining to automotive applications. ► Electrical conductivity and surface tension do not determine fully the spray behavior. ► Non-dimensional analysis was performed to classify flow regimes for each blend. ► We numbers revealed the possibility of droplet secondary break-up. - Abstract: Electrostatically assisted sprays of three fuel blends of bio-butanol, ethanol and heptane were studied experimentally. Mixture composition was selected such that electrical conductivity and surface tension were kept constant for all three mixtures. In this manner, the effect of fuel composition was investigated in a context that broadens the classical focus on the effective decrease of surface tension through the action of electrostatic fields. High-speed visualization was used in order to capture e-spray morphology. In addition, probability density functions of the e-spray droplet size and velocity were measured using Phase-Doppler Anemometry for a variety of flow rates and applied voltages. The dependence of droplet average diameter on both flow rate and applied electric field was highlighted. Polydisperse sprays were observed which was rationalized through the calculation of droplet Weber numbers that pointed to the possibility of a secondary droplet break-up.

Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

International Nuclear Information System (INIS)

Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

1976-01-01

A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

Microstructure and elemental distribution of americium containing MOX fuel under the short term irradiation tests

International Nuclear Information System (INIS)

Tanaka, Kosuke; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin Ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

2008-01-01

In order to investigate the effect of americium addition to MOX fuels on the irradiation behavior, the 'Am-1' program is being conducted in JAEA. The Am-1 program consists of two short term irradiation tests of 10-minute and 24 hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. The PIEs for Am-containing MOX fuels focused on the microstructural evolution and redistribution behavior of Am at the initial stage of irradiation and the results to date are reported

Fuel containment and damage tolerance in large composite primary aircraft structures. Phase 2: Testing

Science.gov (United States)

Sandifer, J. P.; Denny, A.; Wood, M. A.

1985-01-01

Technical issues associated with fuel containment and damage tolerance of composite wing structures for transport aircraft were investigated. Material evaluation tests were conducted on two toughened resin composites: Celion/HX1504 and Celion/5245. These consisted of impact, tension, compression, edge delamination, and double cantilever beam tests. Another test series was conducted on graphite/epoxy box beams simulating a wing cover to spar cap joint configuration of a pressurized fuel tank. These tests evaluated the effectiveness of sealing methods with various fastener types and spacings under fatigue loading and with pressurized fuel. Another test series evaluated the ability of the selected coatings, film, and materials to prevent fuel leakage through 32-ply AS4/2220-1 laminates at various impact energy levels. To verify the structural integrity of the technology demonstration article structural details, tests were conducted on blade stiffened panels and sections. Compression tests were performed on undamaged and impacted stiffened AS4/2220-1 panels and smaller element tests to evaluate stiffener pull-off, side load and failsafe properties. Compression tests were also performed on panels subjected to Zone 2 lightning strikes. All of these data were integrated into a demonstration article representing a moderately loaded area of a transport wing. This test combined lightning strike, pressurized fuel, impact, impact repair, fatigue and residual strength.

Design Package for Fuel Retrieval System Fuel Handling Tool Modification

International Nuclear Information System (INIS)

TEDESCHI, D.J.

2000-01-01

This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports

Fabrication of metallic channel-containing UO2 fuels

International Nuclear Information System (INIS)

Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

2004-01-01

The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

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)

The disposal of Canada's nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock volume 1: summary

International Nuclear Information System (INIS)

Wikjord, A.G.; Baumgartner, P.; Johnson, L.H.; Stanchell, F.W.; Zach, R.; Goodwin, B.W.

1996-06-01

The concept for disposal of Canada's nuclear fuel waste involves isolating the waste in corrosion-resistant containers emplaced and sealed within a vault at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The case for the acceptability of the concept as a means of safely disposing of Canada's nuclear fuel waste is presented in an Environmental Impact Statement (EIS) The disposal concept permits a choice of methods, materials, site locations and designs. The EIS presents a case study of the long-term (i.e., postclosure) performance of a hypothetical implementation of the concept, referred to in this report as the reference disposal system. The reference disposal system is based on borehole emplacement of used CANDU fuel in Grade-2 titanium alloy containers in low-permeability, sparsely fractured plutonic rock of the Canadian Shield. We evaluate the long-term performance of another hypothetical implementation of the concept based on in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. The geological characteristics of the geosphere assumed for this study result in short groundwater travel times from the disposal vault to the surface. In the present study, the principal barrier to the movement of contaminants is the long-lasting copper container. We show that the long-lasting container can effectively compensate for a permeable host rock which results in an unfavourable groundwater flow condition. These studies illustrate the flexibility of AECL's disposal concept to take advantage of the retention, delay, dispersion, dilution and radioactive decay of contaminants in a system of natural barriers provided by the geosphere and hydrosphere and of engineered barriers provided by the waste form, container, buffer, backfills, other vault seals and grouts. In an actual implementation, the engineered system would be designed for the geological conditions encountered at the host site. 34 refs., 2 tabs., 11 figs

Integral-fuel blocks

International Nuclear Information System (INIS)

Cunningham, C.; Simpkin, S.D.

1975-01-01

A prismatic moderator block is described which has fuel-containing channels and coolant channels disposed parallel to each other and to edge faces of the block. The coolant channels are arranged in rows on an equilateral triangular lattice pattern and the fuel-containing channels are disposed in a regular lattice pattern with one fuel-containing channel between and equidistant from each of the coolant channels in each group of three mutually adjacent coolant channels. The edge faces of the block are parallel to the rows of coolant channels and the channels nearest to each edge face are disposed in two rows parallel thereto, with one of the rows containing only coolant channels and the other row containing only fuel-containing channels. (Official Gazette)

FAILED FUEL DISPOSITION STUDY

International Nuclear Information System (INIS)

THIELGES, J.R.

2004-01-01

In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

FAILED FUEL DISPOSITION STUDY

Energy Technology Data Exchange (ETDEWEB)

THIELGES, J.R.

2004-12-20

In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

Probabilistic analysis of the efficiency of the damping devices against nuclear fuel container falling

Science.gov (United States)

Králik, Juraj

2017-07-01

The paper presents the probabilistic and sensitivity analysis of the efficiency of the damping devices cover of nuclear power plant under impact of the container of nuclear fuel of type TK C30 drop. The finite element idealization of nuclear power plant structure is used in space. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall. The experimental results of the shock-damper basic element behavior under impact loads are presented. The Newmark integration method is used for solution of the dynamic equations. The sensitivity and probabilistic analysis of damping devices was realized in the AntHILL and ANSYS software.

Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

International Nuclear Information System (INIS)

Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

2007-01-01

The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

Method of transporting fuel assemblies

International Nuclear Information System (INIS)

Okada, Katsutoshi.

1979-01-01

Purpose: To enable safety transportation of fuel assemblies for FBR type reactors by surrounding each of fuel elements in a wrapper tube by a rubbery, hollow cylindrical container and by sealing medium such as air to the inside of the container. Method: A fuel element is contained in a hollow cylindrical rubber-like tube. The fuel element has an upper end plug, a lower end plug and a wire spirally wound around the outer periphery. Upon transportation of the fuel assemblies, each of the fuel elements is covered with the container and arranged in the wrapper tube and then the fuel assemblies are assembled. Then, medium such as air is sealed for each of the fuel elements by way of an opening and then the opening is tightly closed. Before loading the transported fuel assemblies in the reactor, the medium is discharged through the opening and the container is completely extracted and removed from the inside of the wrapper tube. (Seki, T.)

Fuel transfer system upender using translation drive

International Nuclear Information System (INIS)

Hardin, R.T.

1985-01-01

A transfer system for a nuclear fuel container within a nuclear reactor facility includes a transport car for transporting the fuel container through a transfer tube between a reactor containment handling pool and a spent storage pool. The system includes mechanisms for automatically pivoting the fuel container from its horizontal transport mode to its vertical, fuel loading-unloading mode when the fuel container enters one of the pools. The pivot mechanisms include slotted brackets mounted upon the fuel container, and pivotable pick-up bars for engaging the brackets of the fuel container. As the transport car moves past the pick-up bars, the brackets of the fuel container engage the bars whereby the latter pivot so as to in turn cause pivoting of the fuel container through means of trunnions. Reverse movement of the transport car causes reverse pivoting of the container from the vertical to the horizontal mode and ultimate disengagement of the brackets from the pick-up bars. (author)

Nuclear fuel element

International Nuclear Information System (INIS)

Penrose, R.T.; Thompson, J.R.

1976-01-01

A method of protecting the cladding of a nuclear fuel element from internal attack and a nuclear fuel element for use in the core of a nuclear reactor are disclosed. The nuclear fuel element has disposed therein an additive of a barium-containing material and the barium-containing material collects reactive gases through chemical reaction or adsorption at temperatures ranging from room temperature up to fuel element plenum temperatures. The additive is located in the plenum of the fuel element and preferably in the form of particles in a hollow container having a multiplicity of gas permeable openings in one portion of the container with the openings being of a size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles. The additive is comprised of elemental barium or a barium alloy containing one or more metals in addition to barium such as aluminum, zirconium, nickel, titanium and combinations thereof. 6 claims, 3 drawing figures

Microstructural evolution and Am migration behaviour in Am-containing fuels at the initial stage of irradiation

International Nuclear Information System (INIS)

Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Osaka, Masahiko; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin-ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

2010-01-01

In order to investigate the effect of americium addition to MOX fuels on the irradiation behaviour, the 'Am-1' programme is being conducted in JAEA. The Am-1 programme consists of two short-term irradiation tests of 10-minute and 24-hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post-irradiation examinations (PIE) are in progress. The PIE for Am-containing MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation and the results to date are reported. The successful development of fabrication technology with remote handling and the evaluation of thermo-chemical properties based on the out-of-pile experiments are described with an emphasis on the effects of Am addition on the MOX fuel properties. (authors)

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

1990-01-01

The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

Design of containment system of nuclear fuel attacked by corrosion with leaking fission products

International Nuclear Information System (INIS)

Poblete Maturana, Tomas

2015-01-01

The following report presents the design of an innovative confinement system for the nuclear fuel attacked by corrosion, with leakage of fission products to be used in the RECH-1 nuclear experimental reactor of the Chilean Nuclear Energy Commission, is currently within the framework of the international nuclear waste management program developed by the member countries of the IAEA, including Chile. The main objective of this project is the development of a system that is capable of containing, in the smallest possible volume, the fission products that are released to the reactor coolant medium from the nuclear fuel that are attacked by corrosion. Among the tasks carried out for the development of the project are: the compilation of the necessary bibliography for the selection of the most suitable technology for the retention of the fission products, the calculation of the most important parameters to ensure that the system will operate within ranges that do not compromise the radiological safety, and the design of the hydraulic circuit of the system. The results obtained from the calculations showed that the fuel element confinement system is stable from a thermal point of view since the refrigerant does not under any circumstances reach the saturation temperature and, in addition, from a hydraulic point of view, since the rate at which the refrigerant flows through the hydraulic circuit is low enough so that the deformation of the fuel plates forming the nuclear fuel does not occur. The most appropriate technology for the extraction of fission products according to the literature consulted is by ion exchange. The calculations developed showed that with a very small volume of resins, it is possible to capture all of the non-volatile fission products of a nuclear fuel

A method and apparatus for the manufacture of glass microspheres adapted to contain a thermonuclear fuel

International Nuclear Information System (INIS)

Budrick, R.G.; Nolen, R.L. Jr.; Solomon, D.E.; King, F.T.

1975-01-01

The invention relates to the manufacture of glass microspheres. It refers to a method according to which a sintered glass-powder, whose particles are calibrated, is introduced into a blow-pipe adapted to project said glass-powder particles into a heated flue, said sintered glass-powder containing a pore-forming agent adapted to expand the glass particles into microspheres which are collected in a chamber situated abode said flue. The method can be applied to the manufacture of microspheres adapted to contain a thermonuclear fuel [fr

Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Timothy A. Hyde

2012-06-01

The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

International Nuclear Information System (INIS)

Lerner, Ana Maria; Madariaga, Marcelo

2002-01-01

The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

Green energy: Water-containing acetone–butanol–ethanol diesel blends fueled in diesel engines

International Nuclear Information System (INIS)

Chang, Yu-Cheng; Lee, Wen-Jhy; Lin, Sheng-Lun; Wang, Lin-Chi

2013-01-01

Highlights: • Water-containing ABE solution (W-ABE) in the diesel is a stable fuel blends. • W-ABE can enhance the energy efficiency of diesel engine and act as a green energy. • W-ABE can reduce the PM, NOx, and PAH emissions very significantly. • The W-ABE can be manufactured from waste bio-mass without competition with food. • The W-ABE can be produced without dehydration process and no surfactant addition. - Abstract: Acetone–Butanol–Ethanol (ABE) is considered a “green” energy resource because it emits less carbon than many other fuels and is produced from biomass that is non-edible. To simulate the use of ABE fermentation products without dehydration and no addition of surfactants, a series of water-containing ABE-diesel blends were investigated. By integrating the diesel engine generator (DEG) and diesel engine dynamometer (DED) results, it was found that a diesel emulsion with 20 vol.% ABE-solution and 0.5 vol.% water (ABE20W0.5) enhanced the brake thermal efficiencies (BTE) by 3.26–8.56%. In addition, the emissions of particulate matter (PM), nitrogen oxides (NOx), polycyclic aromatic hydrocarbons (PAHs), and the toxicity equivalency of PAHs (BaP eq ) were reduced by 5.82–61.6%, 3.69–16.4%, 0.699–31.1%, and 2.58–40.2%, respectively, when compared to regular diesel. These benefits resulted from micro-explosion mechanisms, which were caused by water-in-oil droplets, the greater ABE oxygen content, and the cooling effect that is caused by the high vaporization heat of water-containing ABE. Consequently, ABE20W0.5, which is produced by environmentally benign processes (without dehydration and no addition of surfactants), can be a good alternative to diesel because it can improve energy efficiency and reduce pollutant emissions

The disposal of Canada's nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. Volume 2: vault model

International Nuclear Information System (INIS)

Johnson, L.H.; LeNeveu, D.M.; King, F.; Shoesmith, D.W.; Kolar, M.; Oscarson, D.W.; Sunder, S.; Onofrei, C.; Crosthwaite, J.L.

1996-06-01

A study has been undertaken to evaluate the design and long-term performance of a nuclear fuel waste disposal vault based on a concept of in-room emplacement of copper containers at a depth of 500 m in plutonic rock in the Canadian Shield. The containers, each with 72 used CANDU fuel bundles, would be surrounded by clay-based buffer and backfill materials in an array of parallel rooms, with the excavation boundary assumed to have an excavation-disturbed zone (EDZ) with a higher permeability than the surrounding rock. In the anoxic conditions of deep rock of the Canadian Shield, the copper containers are expected to survive for >10 6 a. Thus container manufacturing defects, which are assumed to affect approximately 1 in 5000 containers, would be the only potential source of radionuclide release in the vault. The vault model is a computer code that simulates the release of radionuclides that would occur upon contact of the used fuel with groundwater, the diffusive transport of these radionuclides through the defect in the container shell and the surrounding buffer, and their dispersive and convective transport through the backfill and EDZ into the surrounding rock. The vault model uses a computationally efficient boundary integral model (BIM) that simulates radionuclide mass transport in the engineered barrier system as a point source (representing the defective container) that releases radionuclides into concentric cylinders, that represent the buffer, backfill and EDZ. A 3-dimensional finite-element model is used to verify the accuracy of the BIM. The results obtained in the present study indicates the effectiveness of a design using in-room emplacement of long-lived containers in providing a safe disposal system even under permeable geosphere conditions. (author). refs., tabs., figs

The disposal of Canada`s nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock volume 1: summary

Energy Technology Data Exchange (ETDEWEB)

Wikjord, A G; Baumgartner, P; Johnson, L H; Stanchell, F W; Zach, R; Goodwin, B W

1996-06-01

The concept for disposal of Canada`s nuclear fuel waste involves isolating the waste in corrosion-resistant containers emplaced and sealed within a vault at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The case for the acceptability of the concept as a means of safely disposing of Canada`s nuclear fuel waste is presented in an Environmental Impact Statement (EIS) The disposal concept permits a choice of methods, materials, site locations and designs. The EIS presents a case study of the long-term (i.e., postclosure) performance of a hypothetical implementation of the concept, referred to in this report as the reference disposal system. The reference disposal system is based on borehole emplacement of used CANDU fuel in Grade-2 titanium alloy containers in low-permeability, sparsely fractured plutonic rock of the Canadian Shield. We evaluate the long-term performance of another hypothetical implementation of the concept based on in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. The geological characteristics of the geosphere assumed for this study result in short groundwater travel times from the disposal vault to the surface. In the present study, the principal barrier to the movement of contaminants is the long-lasting copper container. We show that the long-lasting container can effectively compensate for a permeable host rock which results in an unfavourable groundwater flow condition. These studies illustrate the flexibility of AECL`s disposal concept to take advantage of the retention, delay, dispersion, dilution and radioactive decay of contaminants in a system of natural barriers provided by the geosphere and hydrosphere and of engineered barriers provided by the waste form, container, buffer, backfills, other vault seals and grouts. In an actual implementation, the engineered system would be designed for the geological conditions encountered at the host site. 34 refs., 2 tabs., 11 figs.

Storage container for radioactive fuel elements

International Nuclear Information System (INIS)

1984-01-01

The interim storage cask for spent fuel elements or the glass moulds for high-level radioactive waste are made up of heat-resistant, reinforced concrete with chambers and highgrade steel lining. Cooling systems with natural air circulation are connected with the chambers. (HP) [de

Spent nuclear fuel storage device and spent nuclear fuel storage method using the device

International Nuclear Information System (INIS)

Tani, Yutaro

1998-01-01

Storage cells attachably/detachably support nuclear fuel containing vessels while keeping the vertical posture of them. A ventilation pipe which forms air channels for ventilating air to the outer circumference of the nuclear fuel containing vessel is disposed at the outer circumference of the nuclear fuel containing vessel contained in the storage cell. A shielding port for keeping the support openings gas tightly is moved, and a communication port thereof can be aligned with the upper portion of the support opening. The lower end of the transporting and containing vessel is placed on the shielding port, and an opening/closing shutter is opened. The gas tightness is kept by the shielding port, the nuclear fuel containing vessel filled with spent nuclear fuels is inserted to the support opening and supported. Then, the support opening is closed by a sealing lid. (I.N.)

Fuel assembly

International Nuclear Information System (INIS)

Watanabe, Shoichi; Hirano, Yasushi.

1998-01-01

A one-half or more of entire fuel rods in a fuel assembly comprises MOX fuel rods containing less than 1wt% of burnable poisons, and at least a portion of the burnable poisons comprises gadolinium. Then, surplus reactivity at an initial stage of operation cycle is controlled to eliminate burnable poisons remained unburnt at a final stage, as well as increase thermal reactivity. In addition, the content of fission plutonium is determined to greater than the content of uranium 235, and fuel rods at corner portions are made not to incorporate burnable poisons. Fuel rods not containing burnable poisons are disposed at positions in adjacent with fuel rods facing to a water rod at one or two directions. Local power at radial center of the fuel assembly is increased to flatten the distortion of radial power distribution. (N.H.)

Process for surface treatment of zirconium-containing cladding materials for fuel element or other components for nuclear reactors

International Nuclear Information System (INIS)

Videm, K.G.; Lunde, L.R.; Kooyman, H.H.

1975-01-01

A process for the surface treatment of zirconium-base cladding materials for fuel elements or other components for nuclear reactors is described. The treatment includes pickling the cladding material in a fluoride-containing bath, and then applying a protective coating through oxidation to the pickled cladding material. The fluoride-containing contaminants which remain on the surface of the cladding material during pickling are removed or rendered harmless by anodic oxidation

Use of water containing acetone–butanol–ethanol for NOx-PM (nitrogen oxide-particulate matter) trade-off in the diesel engine fueled with biodiesel

International Nuclear Information System (INIS)

Chang, Yu-Cheng; Lee, Wen-Jhy; Wu, Tser Son; Wu, Chang-Yu; Chen, Shui-Jen

2014-01-01

Fuel blends that contain biodiesel are known to produce greater NO x (nitrogen oxide) emissions in diesel engine exhaust than regular diesel, and this is one of the key barriers to the wider adoption of biodiesel as an alternative fuel. In this study, a water-containing ABE (acetone–butanol–ethanol) solution, which simulates products that are produced from biomass fermentation without dehydration processing, was tested as a biodiesel-diesel blend additive to lower NO x emissions from diesel engines. The energy efficiency and the PM (particulate matter) and PAHs (polycyclic aromatic hydrocarbons) emissions were investigated and compared under various operating conditions. Although biodiesel had greater NO x emissions, the blends that contained 25% of the water-containing ABE solution had significantly lower NO x (4.30–30.7%), PM (10.9–63.1%), and PAH (polycyclic aromatic hydrocarbon) emissions (26.7–67.6%) than the biodiesel–diesel blends and regular diesel, respectively. In addition, the energy efficiency of this new blend was 0.372–7.88% higher with respect to both the biodiesel–diesel blends and regular diesel. Because dehydration and surfactant addition are not necessary, the application of ABE–biodiesel–diesel blends can simplify fuel production processes, reduce energy consumption, and lower pollutant emissions, meaning that the ABE–biodiesel–diesel blend is a promising green fuel. - Highlights: • Water-containing ABE (acetone–butanol–ethanol)–biodiesel–diesel was tested in a diesel engine. • The addition of ABE to biodiesel–diesel blends can enhance the energy efficiency. • The addition of ABE can solve the problem of NO x -PM (nitrogen oxide-particulate matter) trade-off when using biodiesel. • PAHs (polycyclic aromatic hydrocarbons) can be further reduced by adding ABE in biodiesel–diesel blends. • Fuel production was simplified due to the acceptance of water in ABE

A fuel response model for the design of spent fuel shipping casks

International Nuclear Information System (INIS)

Malinauskas, A.P.; Duffey, T.A.; Einziger, R.E.; Hobbins, R.R.; Jordon, H.; Rashid, Y.R.; Barrett, P.R.; Sanders, T.L.

1989-01-01

The radiological source terms pertinent to spent fuel shipping cask safety assessments are of three distinct origins. One of these concerns residual contamination within the cask due to handling operations and previous shipments. A second is associated with debris (''crud'') that had been deposited on the fuel rods in the course of reactor operation, and a third involves the radioactive material contained within the rods. Although the lattermost source of radiotoxic material overwhelms the others in terms of inventory, its release into the shipping cask, and thence into the biosphere, requires the breach of an additional release barrier, viz., the fuel rod cladding. Hence, except for the special case involving the transport of fuel rods containing previously breached claddings, considerations of the source terms due to material contained in the fuel rods are complicated by the need to address the likelihood of fuel cladding failure during transport. The purpose of this report is to describe a methodology for estimating the shipping cask source terms contribution due to radioactive material contained within the spent fuel rods. Thus, the probability of fuel cladding failure as well as radioactivity release is addressed. 8 refs., 2 tabs

Improved nuclear fuel element

International Nuclear Information System (INIS)

Klepfer, H.H.

1974-01-01

A nuclear fuel element is described which comprises: 1) an elongated clad container, 2) a layer of high lubricity material being disposed in and adjacent to the clad container, 3) a low neutron capture cross section metal liner being disposed in the clad container and adjacent to the layer, 4) a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, 5) an enclosure integrally secured and sealed at each end of the container, and a nuclear fuel material retaining means positioned in the cavity. (author)

Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

International Nuclear Information System (INIS)

Whitesides, G.H.; Stephens, M.E.

1984-01-01

During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

A copper container corrosion model for the in-room emplacement of used CANDU fuel

International Nuclear Information System (INIS)

King, F.

1996-11-01

Copper containers in a Canadian nuclear fuel waste disposal vault are expected to undergo uniform corrosion and, possibly, pitting. The corrosion behaviour of the containers will be dictated by the evolution of environmental conditions within the disposal vault. The environment will evolve from an early warm, oxidizing phase, during which fast uniform corrosion and pitting may occur, to an indefinite period of cool, anoxic conditions, during which the container will only be susceptible to slow uniform corrosion. The results of corrosion and electrochemical studies of the uniform corrosion of Cu in O 2 -containing Cl - solutions are discussed and a detailed reaction mechanism presented. The relevant literature on pitting corrosion is briefly reviewed and models for the prediction of pit depth discussed. The potential for microbially influenced corrosion and stress-corrosion cracking is discussed, as are vapour-phase corrosion and the effects of β-radiation. The use of natural analogues for justifying long-term corrosion predictions is also considered. Finally, a model for uniform corrosion and pitting is presented and container lifetimes predicted. Copper containers having a minimum wall thickness of 25.4 mm are not predicted to fail by corrosion in periods 6 a. Thus, despite the assumption of poor rock quality made here, the safety of the entire disposal concept can be assured by the use of a long-lived container. (author). 125 refs., 1 tab., 24 figs

Corrosion of copper containers prior to saturation of a nuclear fuel waste disposal vault

International Nuclear Information System (INIS)

King, F.; Kolar, M.

1997-12-01

The buffer material surrounding the containers in a Canadian nuclear fuel waste disposal vault will partially desiccate as a result of the elevated temperature at the container surface. This will lead to a period of corrosion in a moist air atmosphere. Corrosion will either take the form of slow oxidation if the container surface remains dry or aqueous electrochemical corrosion if the surface is wetted by a thin liquid film. The relevant literature is reviewed, from which it is concluded that corrosion should be uniform in nature, except if the surface is wetted, in which case localized corrosion is a possibility. A quantitative analysis of the extent and rate of uniform corrosion during the unsaturated period is presented. Two bounding cases are considered: first, the case of slow oxidation in moist air following either logarithmic or parabolic oxide-growth kinetics and, second, the case of electrochemically based corrosion occurring in a thin liquid film uninhibited by the growth of corrosion products. (author)

Ammonia as a Suitable Fuel for Fuel Cells

International Nuclear Information System (INIS)

Lan, Rong; Tao, Shanwen

2014-01-01

Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5 wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

SPHERE: Irradiation of sphere-pac fuel of UPuO2−x containing 3% Americium

International Nuclear Information System (INIS)

D’Agata, E.; Hania, P.R.; McGinley, J.; Somers, J.; Sciolla, C.; Baas, P.J.; Kamer, S.; Okel, R.A.F.; Bobeldijk, I.; Delage, F.; Bejaoui, S.

2014-01-01

Highlights: • SPHERE is designed to check the behaviour of MADF sphere-pac concept. • MADF sphere-pac are compared with MADF pellet. • Swelling, helium release and restructuring behaviour will be the main output of the experiment. • An experiment to check sphere-pac MABB fuel behaviour is now under design. - Abstract: Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like 241 Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. The SPHERE irradiation experiment is the latest of a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS) performed in the HFR (High Flux Reactor). The SPHERE experiment is carried out in the framework of the 4-year project FAIRFUELS of the EURATOM 7th Framework Programme (FP7). During the past years of experimental works in the field of transmutation and tests of innovative nuclear fuels, the release or trapping of helium as well as helium induced fuel swelling have been shown to be the key issues for the design of Am-bearing targets. The main objective of the SPHERE experiment is to study the in-pile behaviour of fuel containing 3% of americium and to compare the behaviour of sphere-pac fuel to pellet fuel, in particular the role of microstructure and temperature on fission gas release (mainly He) and on fuel swelling. The SPHERE experiment is being irradiated since September 2013 in the HFR in Petten (The Netherlands) and is expected to be terminated in spring 2015. The experiment has been designed to last up to 18 reactor cycles (corresponding to 18 months) but may reach its target earlier. This paper discusses the rationale and objective of the SPHERE experiment and provides a general description of its design

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

International Nuclear Information System (INIS)

Nordin, A.

1992-12-01

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

Proportioning equipment for vibration filling and compacting of grain materials in pipe containers, especially of fuel elements

International Nuclear Information System (INIS)

Pinkas, V.; Filip, Z.; Beranek, J.

1981-01-01

The equipment consists of a base plate to which are attached the fastening collar fo the pipe container and the guide column with the height-adjustable support. The filling pipe is fixed to the support. The proportioning equipment prevents particles of grain material from segregation, thus allowing to achieve homogeneity of the material in the whole volume to be compacted. It also allows determining the height of the column of material in the pipe container without destructive effects on the stacked material. The equipment is designed for the manufacture of shortened fuel elements. (J.B.)

Ammonia as a suitable fuel for fuel cells

Directory of Open Access Journals (Sweden)

Rong eLan

2014-08-01

Full Text Available Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

Research on treatment of wastewater containing heavy metal by microbial fuel cell

Science.gov (United States)

Chen, Zixuan; Lu, Xun; Yin, Ruixia; Luo, Yunyi; Mai, Hanjian; Zhang, Nan; Xiong, Jingfang; Zhang, Hongguo; Tang, Jinfeng; Luo, Dinggui

2018-02-01

With rapid development of social economy, serious problem has been caused by wastewater containing heavy metals, which was difficult to be treated by many kinds of traditional treatment methods, such as complex processes, high cost or easy to cause secondary pollution. As a novel biological treatment technology, microbial fuel cells (MFC) can generate electric energy while dealing with wastewater, which was proposed and extensively studied. This paper introduced the working principle of MFC, the classification of cathode, and the research progress on the treatment of wastewater containing Cr(VI), Cu(II), Ag(I), Mn(II) and Cd(II) by MFC. The study found that different cathode, different heavy metals anddifferent hybrid systems would affect the performance of the system and removal effect for heavy metal in MFC. MFC was a highly potential pollution control technology. Until now, the research was still in the laboratory stage. Its industrial application for recovery of heavy metal ion, improving the energy recovery rate and improvement or innovation of system were worthy of further research.

Stationary Liquid Fuel Fast Reactor

International Nuclear Information System (INIS)

Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

2015-01-01

For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

Stationary Liquid Fuel Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-09-30

For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

Energy Technology Data Exchange (ETDEWEB)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-02-26

of MOX-fuel production waste is incineration or calcination, alkali sintering, and dissolution of sintered products in nitric acid. Insoluble residues are then mixed with vitrifying components and Pu sludges, vitrified, and sent for storage and disposal. Implementation of the intergovernmental agreement between Russia and the United States (US) regarding the utilization of 34 tons of weapons plutonium will also require treatment of Pu containing MOX fabrication wastes at the MCC radiochemical production plant.

Magnetic signature surveillance of nuclear fuel

International Nuclear Information System (INIS)

Bernatowicz, H.; Schoenig, F.C.

1981-01-01

Typical nuclear fuel material contains tramp ferromagnetic particles of random size and distribution. Also, selected amounts of paramagnetic or ferromagnetic material can be added at random or at known positions in the fuel material. The fuel material in its non-magnetic container is scanned along its length by magnetic susceptibility detecting apparatus whereby susceptibility changes along its length are obtained and provide a unique signal waveform of the container of fuel material as a signature thereof. The output signature is stored. At subsequent times in its life the container is again scanned and respective signatures obtained which are compared with the initially obtained signature, any differences indicating alteration or tampering with the fuel material. If the fuel material includes a paramagnetic additive by taking two measurements along the container the effects thereof can be cancelled out. (author)

Device for manipulating a nuclear reactor fuel element in a fuel element pond containing water

International Nuclear Information System (INIS)

Jabsen, F.S.

1977-01-01

Using this device a fuel element can be manipulated inside a water filled storage pond for inspection purposes. A transport arrangement which is normally situated above such a pond is modified for this purpose. A crane bridge runs on rails on the upper edge of the pond. A type of trolley runs transversely to the direction of travel of the bridge between 2 wide flange supports forming the crane support. During movement this trolley moves a submerged combination of periscope and TV camera pendant from it at about half the pond height horizontally along the crane support. 2 vehicles move between these on 4 rollers each, on the under flanges of the crane support at spacings of about one fuel element length. A pendant arm of the same length as the periscope dips vertically into the pond from each vehicle. There is a bar of about fuel element length resting on the lower ends of both arms. The surface of a fuel element lying on this bar can be inspected through the periscope on longitudinal travel of the trolley. The bar with the fuel element can be rotated 90 0 downwards into a vertical position after removal of one or more rotating kingpins and release of a rope hanging on the end away from the kingpin. The rope is actuated by a winch on the crane support. The bar has vertical plates at both ends to hold the fuel element in its vertical position. (HP) [de

Facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Hennings, U.

1987-01-01

Patent for facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies, which are arranged within a building in a horizontal position and are cooled by a gas stream, whereby the building has a storage and a loading zone, characterized by the fact that pallet trucks arranged one above the other in a row and such that an interspace is left for the receiving positions for the containers, the the pallet trucks can be moved along rails that extend between two side walls arranged opposite to one another in the storage zone, that the storage zone can be loaded and unloaded by opening located in these two side walls, and that the gas stream only circulates within the building

Plutonium-containing aerosols found within containment enclosures in industrial mixed-oxide reactor fuel fabrication

International Nuclear Information System (INIS)

Newton, G.J.; Yeh, H.C.; Stanley, J.A.

1977-01-01

Mixed oxide (PuO 2 and UO 2 ) nuclear reactor fuel pellets are fabricated within safety enclosures at Babcock and Wilcox's Park Township site near Apollo, PA. Forty-two sample runs of plutonium-containing aerosols were taken from within glove boxes during routine industrial operations. A small, seven-stage cascade impactor and the Lovelace Aerosol Particle Separator (LAPS) were used to determine aerodynamic size distribution and gross alpha aerosol concentration. Powder comminution and blending produced aerosols with lognormal size distributions characterized by activity median aerodynamic diameters (AMAD) of 1.89 +- 0.33 μm, sigma/sub g/ = 1.62 +- 0.09 and a gross alpha aerosol concentration range of 0.1 to 150 nCi/l. Slug pressing and grinding produced aerosols of AMAD = 3.08 +- 0.1 μm, sigma/sub g/ = 1.53 +- 0.01 and AMAD = 2.26 +- 0.16 μm, sigma/sub g/ = 1.68 +- 0.20, respectively. Gross alpha aerosol concentrations ranged from 3.4 to 450 nCi/l. Centerless grinding produced similar-sized aerosols but the gross alpha concentration ranged from 220 to 1690 nCi/l. In vitro solubility studies on selected LAPS samples in a lung fluid simulant indicate that plutonium mixed-oxide aerosols are more soluble than laboratory-produced plutonium aerosols

A means to a cleaner environment: energy efficient platinum-containing fuel cells to be introduced commercially in the early 1990s

Energy Technology Data Exchange (ETDEWEB)

Anon,

1991-01-01

The twelfth National Fuel Cell Seminar held in Phoenix, Arizona, U.S.A. from 26th to 28th November 1990, was attended by 450 delegates from 17 countries, representing both developers and potential users worldwide. ''Fuel Cells - An Answer to a Cleaner Environment'' was the key theme running throughout the conference. This was strongly linked to the firm belief that fuel cells would make an important contribution to the world's energy needs over the next ten years as economically viable fuel cell power plants become commercialised. The conference heard that platinum containing phosphoric acid fuel cells (PAFC) are soon to be produced on a commercial basis. Both Fuji and Toshiba announced that they had opened PAFC production facilities in Japan during 1990. The key progress regarding development of Proton Exchange Membrane Fuel Cells (PEMFC) was the announcement of a U.S. Government sponsored programme, to be led to General Motors, to produce a PEMFC powered motor vehicle. The conference reflected the growing extent of the multi-national collaborations that are now underway to develop fuel cell technologies. (author).

Predicting the effects of microbial activity on the corrosion of copper nuclear fuel waste disposal containers

International Nuclear Information System (INIS)

King, F.; Stroes-Gascoyne, S.

1996-08-01

Microbially influenced corrosion (MIC) of copper nuclear fuel waste containers may occur in a disposal vault located 500-1000 m underground in the granitic rock of the Canadian Shield. The extent and diversity of microbial activity in the vault is expected to be limited initially because of the aggressive conditions produced by γ-radiation, elevated temperatures and desiccation of the clay-based buffer in which the containers will be embedded. Experimental results on the heat- and radiation-sensitivity of the natural microbiota in buffer material are presented. The data suggest that the low water activity in the buffer material will severely limit the growth of microbes near the container. The most likely form of MIC involves sulphate-reducing bacteria (SRB). Electrochemical experiments using a clay-covered copper electrode have shown that sulphide ions produced by SRB could diffuse through buffer material and induce corrosion of the container. A method to predict the long-term corrosion behaviour is presented. (author)

Optimization strategies for cask design and container loading in long term spent fuel storage

International Nuclear Information System (INIS)

2006-12-01

As delays are incurred in implementing reprocessing and in planning for geologic repositories, storage of increasing quantities of spent fuel for extended durations is becoming a growing reality. Accordingly, effective management of spent fuel continues to be a priority topic. In response, the IAEA has organized a series of meetings to identify cask loading optimisation issues in preparation for a technical publication on Optimization Strategies for Cask/Container Loading in Long Term Spent Fuel Storage. This publication outlines the optimisation process for cask design, licensing and utilization, describing three principal groups of optimization activities in terms of relevant technical considerations such as criticality, shielding, structural design, operations, maintenance and retrievability. The optimization process for cask design, licensing, and utilization is outlined. The general objectives for the design of storage casks, including storage casks that are intended to be transportable, are summarized. The nature of optimization within the design process is described. The typical regulatory and licensing process is outlined, focusing on the roles of safety regulations, the regulator, and the designer/applicant in the optimization process. Based on the foregoing, a description of the three principal groups of optimization activities is provided. The subsequent chapters of this document then describe the specific optimization activities within these three activity groups, in each of the several design disciplines

Status of irradiation testing and PIE of MOX (Pu-containing) fuel

International Nuclear Information System (INIS)

Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

1995-01-01

This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

Quality of water from the pool, original containers and aluminum drums used for storage of spent nuclear fuel

International Nuclear Information System (INIS)

Idjakovic, Z.; Milonjic, S.; Cupic, S.

2001-01-01

Results of chemical analyses of water from the pool, including original containers and aluminium drums, for storage of spent nuclear fuel of the research reactor RA at the VINCA Institute and a short survey of the water properties from similar pools of other countries are presented in the paper. (author)

Castor oil polyurethane as a coating option for spent nuclear fuel disposal containment

Energy Technology Data Exchange (ETDEWEB)

Mortley, A.; Bonin, H.W.; Bui, V.T. [Dept. of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario (Canada)

2009-07-01

Castor oil polyurethane (COPU) coatings are being proposed as an additional barrier in the design of the copper containers to store spent nuclear fuel in Canada. The present work investigates the variation in the physicomechanical properties of two COPUs, based on an aliphatic and aromatic diisocyanate, as a function of ionizing radiation dose and dose rate. The changes in physicomechanical properties have shown that radiation, regardless of dose rate and isocyanate structure, increases the values of the modulus and the ultimate tensile strength when compared with those of the unirradiated samples, with aromatic based polyurethanes being more susceptible to variation than aliphatic based ones. (author)

Castor oil polyurethane as a coating option for spent nuclear fuel disposal containment

International Nuclear Information System (INIS)

Mortley, A.; Bonin, H.W.; Bui, V.T.

2009-01-01

Castor oil polyurethane (COPU) coatings are being proposed as an additional barrier in the design of the copper containers to store spent nuclear fuel in Canada. The present work investigates the variation in the physicomechanical properties of two COPUs, based on an aliphatic and aromatic diisocyanate, as a function of ionizing radiation dose and dose rate. The changes in physicomechanical properties have shown that radiation, regardless of dose rate and isocyanate structure, increases the values of the modulus and the ultimate tensile strength when compared with those of the unirradiated samples, with aromatic based polyurethanes being more susceptible to variation than aliphatic based ones. (author)

Mechanistic modelling of the corrosion behaviour of copper nuclear fuel waste containers

Energy Technology Data Exchange (ETDEWEB)

King, F; Kolar, M

1996-10-01

A mechanistic model has been developed to predict the long-term corrosion behaviour of copper nuclear fuel waste containers in a Canadian disposal vault. The model is based on a detailed description of the electrochemical, chemical, adsorption and mass-transport processes involved in the uniform corrosion of copper, developed from the results of an extensive experimental program. Predictions from the model are compared with the results of some of these experiments and with observations from a bronze cannon submerged in seawater saturated clay sediments. Quantitative comparisons are made between the observed and predicted corrosion potential, corrosion rate and copper concentration profiles adjacent to the corroding surface, as a way of validating the long-term model predictions. (author). 12 refs., 5 figs.

Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

Energy Technology Data Exchange (ETDEWEB)

Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

Spent fuel dry storage technology development: thermal evaluation of three adjacent drywells (each containing a 0.6 kW PWR spent fuel assembly)

International Nuclear Information System (INIS)

Unterzuber, R.; Hanson, J.P.

1981-09-01

A spent fuel Adjacent Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing three nearly identical pressurized water reactor spent fuel assemblies each having a decay heat level of approximately 0.6 kW. Each fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell storage cell for thermal testing. Each fuel assembly was sealed inside a 14-in. diam, 168-in.-long stainless steel canister and attached to a concrete-filled, 20-in.-diam, 34-in.-long, shield plug. The canister assembly was then placed in a carbon steel drywell liner which had been grouted into a hole drilled in the soil adjacent to E-MAD. The three drywells were located 25 feet apart in a linear array. Thermocouples, provided to measure canister, liner and soil temperatures, were inserted into tubes on the outside of the canister and drywell liner and were attached to plastic pipes which were grouted into holes in the soil. Temperatures from the three drywells and the adjacent soil were recorded throughout the Adjacent Drywell Test. Drywell thermal data showed virtually no thermal interaction between adjacent drywells. However, peak temperatures reached by the three drywells did show a fairly significant difference. Peak canister and drywell liner temperatures were reached in August 1981 for all three drywells. The two previously unused drywells responded similarly with peak canister and liner temperatures reaching 199 0 F and 158 0 F, respectively. Comparable peak temperatures for the third drywell which had previously contained spent fuel for nearly 21 months prior to the Adjacent Drywell Test reached 210 0 F for the canister and 169 0 F for the drywell liner. This difference is attributed to a decrease in soil thermal conductivity caused by the dryout of soil around the drywell used for previous spent fuel testing

Reactive-transport model for the prediction of the uniform corrosion behaviour of copper used fuel containers

International Nuclear Information System (INIS)

King, F.; Kolar, M.; Maak, P.

2008-01-01

Used fuel containers in a deep geological repository will be subject to various forms of corrosion. For containers made from oxygen-free, phosphorus-doped copper, the most likely corrosion processes are uniform corrosion, underdeposit corrosion, stress corrosion cracking, and microbiologically influenced corrosion. The environmental conditions within the repository are expected to evolve with time, changing from warm and oxidizing initially to cool and anoxic in the long-term. In response, the corrosion behaviour of the containers will also change with time as the repository environment evolve. A reactive-transport model has been developed to predict the time-dependent uniform corrosion behaviour of the container. The model is based on an experimentally-based reaction scheme that accounts for the various chemical, microbiological, electrochemical, precipitation/dissolution, adsorption/desorption, redox, and mass-transport processes at the container surface and in the compacted bentonite-based sealing materials within the repository. Coupling of the electrochemical interfacial reactions with processes in the bentonite buffer material allows the effect of the evolution of the repository environment on the corrosion behaviour of the container to be taken into account. The Copper Corrosion Model for Uniform Corrosion predicts the time-dependent corrosion rate and corrosion potential of the container, as well as the evolution of the near-field environment

Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

NARCIS (Netherlands)

Wild, de P.J.; Nyqvist, R.G.; Bruijn, de F.A.; Stobbe, E.R.

2006-01-01

Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based(e.g. proton exchange membrane fuel cells (PEMFC) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To preventdetrimental effects on the

Shipping container for nuclear fuels

International Nuclear Information System (INIS)

Housholder, W.R.; Greer, N.L.

1976-01-01

A container for nuclear materials is described wherein a specially and uniquely constructed pressure vessel and gamma shield assembly for holding the nuclear materials is provided in a housing, and wherein a positioning means extends between the housing and the assembly for spacing the same, insulation in the housing essentially filling the space between the assembly and housing, the insulation comprising beads, globules or the like of water encapsulated in plastic and which, in one important embodiment, contains neutron absorbing matter

Sulfur- and nitrogen-containing phenol-formaldehyde co-resites for probing the thermal behaviour of heteroatomic forms in solid fuels

Energy Technology Data Exchange (ETDEWEB)

Ismail, K.; Sirkecioglu, O.; Andresen, J.M.; Brown, S.D.; Hall, P.J.; Snape, C.E. [University of Strathclyde, Glasgow (United Kingdom). Dept. of Pure and Applied Chemistry

1996-09-01

In order to probe the formation of sulfur- and nitrogen-containing gases during the pyrolysis and combustion of coals and other solid fuels, non-softening model substrates are required. In this respect phenol-formaldehyde (PF) resins are ideal since they readily facilitate the incorporation of individual heteroatomic functions into a highly crosslinked matrix. A series of sulfur- and nitrogen-containing co-resites were prepared using phenol with, as the second component, thiophene, dibenzothiophene, diphenylsulfide, benzyl phenyl sulfide, thioanisole, 8-hydroxyquinoline and 2-hydroxycarbazole. A mole ratio of 3:1 (phenol: heteroatom-containing component) was used. Resoles containing diphenyldisulfide were also prepared but, due to the comparable bond strengths of the S-S and C-O linkages, a curing temperature of only 130{degree}C was used to avoid cleavage of the disulfide bond. The virtually complete elimination of ether and methylol functions from the resoles by curing at 200{degree}C was monitored by solid-state {sup 13}C nuclear magnetic resonance spectroscopy. The resultant resites were also characterized by sulfur K-edge X-ray absorption near-edge structure (XANES) spectroscopy, X-ray photoelectron spectroscopy and differential scanning calorimetry. Simple air oxidation was found to selectively convert the aliphatic-bound sulfur to a mixture of sulfones and sulfoxides. Applications of the resites in fuel science are described.

An oxyde mixture fuel containing uranium and plutonium dioxides and process to obtain this oxyde mixture

International Nuclear Information System (INIS)

Hannerz, K.

1976-01-01

An oxide-mixture fuel containing uranium and plutonium dioxides having the slage of spherical, or nearly spherical, oxide-mixture particles with a diameter within the range of from 0.2 to 2 mn charactarized in that each oxide-mixture particles is provided with an outer layer comprising mainly UO2, the thickness of which is at least 0.05; whereas the inner portion of the oxide-mixture particles comprises mainly PUO 2

Fuel Flexible, Low Emission Catalytic Combustor for Opportunity Fuel Applications

Energy Technology Data Exchange (ETDEWEB)

Eteman, Shahrokh

2013-06-30

Limited fuel resources, increasing energy demand and stringent emission regulations are drivers to evaluate process off-gases or process waste streams as fuels for power generation. Often these process waste streams have low energy content and/or highly reactive components. Operability of low energy content fuels in gas turbines leads to issues such as unstable and incomplete combustion. On the other hand, fuels containing higher-order hydrocarbons lead to flashback and auto-ignition issues. Due to above reasons, these fuels cannot be used directly without modifications or efficiency penalties in gas turbine engines. To enable the use of these wide variety of fuels in gas turbine engines a rich catalytic lean burn (RCL®) combustion system was developed and tested in a subscale high pressure (10 atm.) rig. The RCL® injector provided stability and extended turndown to low Btu fuels due to catalytic pre-reaction. Previous work has shown promise with fuels such as blast furnace gas (BFG) with LHV of 85 Btu/ft3 successfully combusted. This program extends on this work by further modifying the combustor to achieve greater catalytic stability enhancement. Fuels containing low energy content such as weak natural gas with a Lower Heating Value (LHV) of 6.5 MJ/m3 (180 Btu/ft3 to natural gas fuels containing higher hydrocarbon (e.g ethane) with LHV of 37.6 MJ/m3 (1010 Btu/ft3) were demonstrated with improved combustion stability; an extended turndown (defined as the difference between catalytic and non-catalytic lean blow out) of greater than 250oF was achieved with CO and NOx emissions lower than 5 ppm corrected to 15% O2. In addition, for highly reactive fuels the catalytic region preferentially pre-reacted the higher order hydrocarbons with no events of flashback or auto-ignition allowing a stable and safe operation with low NOx and CO emissions.

Nuclear fuel element

International Nuclear Information System (INIS)

Grossman, L.N.; Levin, H.A.

1975-01-01

A nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smallr than the size of the particles. The container is preferably held in the spring in the plenum of the fuel element. (E.C.B.)

On the problem of radiation hazard of fuel-containing materials in conversion of object <>; O probleme yadernoj opasnosti toplivosoderzhashchikh materialov pri preobrazovanii ob`ekta << Ukrytie >>.

Energy Technology Data Exchange (ETDEWEB)

Badovskij, V P; Dubar, L V; Shevchenko, S V [Myizhgaluzevij Naukovo-Tekhnyichnij Tsentr ` ` Ukrittya ` ` , Natsyional` na Akademyiya Nauk Ukrayini, Chornobil` (Ukraine)

1994-12-31

A problem of radiation hazard of fuel-containing materials (FCM) in object << Shelter >> is discussed. A conservative approach is applied. The approach was developed with allowance for a conservative estimation of fuel distribution in FCM and errors of the FCM hazard measurements. A version of 4PB accident development is proposed. According to this version there is considerable part of fuel in 305/2 room which forms FCM with average fuel concentration in this room over 25%.

Fuel characteristics pertinent to the design of aircraft fuel systems

Science.gov (United States)

Barnett, Henry C; Hibbard, R R

1953-01-01

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

Verification of the burn-up of spent fuel assemblies by means of the Consulha containment/surveillance system

International Nuclear Information System (INIS)

Daniel, G.; Gourlez, P.

1991-01-01

CONSULHA is a containment/surveillance system which has been developed as part of the French Support Programme for the IAEA Safeguards in cooperation with EURATOM and was designed to meet the IAEA EURATOM requirements for the verification of nuclear materials. This system will make it possible to count movements and verify irradiation of spent fuel assemblies in industrial facilities such as reprocessing plants and nuclear reactors

Arrangement and statistics of storage containers of spent fuel for assemblies of the SFP of NPP-L V, Unit 1

International Nuclear Information System (INIS)

Mijangos D, Z. E.; Vargas A, A. F.; Amador C, C.

2014-10-01

This work presents the determination of assemblies of the spent fuel pool (SFP) of the nuclear power plant of Laguna Verde (NPP-L V) which are candidates to be assigned to storage containers of independent spent fuel, with the objective of liberating decay heat and to have more space in the SFP, for the store of retired assemblies of the reactors in future reloads of NPP-L V, besides that the removed assemblies of the SFP should be stored in specific containers to guarantee the physical safety of them, as well as the radiological protection to the population and the environment. The design of the containers considered in this work is to store a maximum of 69 assemblies; it has a thermal capacity of 26 kilowatts and allows storing assemblies with a minimum of 5 years of have been extracted of the reactor core. Is considered that in 2016 start the storage of the spent assemblies on the containers, the candidates assemblies to store cover from the first reload in 1991, until the assemblies deposited in the SFP in the 14 reload in 2010; therefore in 2016, such assemblies will have fulfilled with the criteria of 5 years of have been removed of the Reactor, also the 69 assemblies assigned to each container will have a resulting decay heat that does not exceed the thermal capacity of the container, but that in great percentage approximates to the same one, and this way to take full advantage of their storage capacity and thermal capacity for each container. This work also contains the arrangement to accommodate the assemblies in the containers; such arrangement is constituted by areas according to the decay heat of each assembly. (Author)

Update on ASME rules for spent nuclear fuel and high level radioactive material and waste storage containments

International Nuclear Information System (INIS)

Ralph S. Hill III; Foster, G.M.

2005-01-01

In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) was published. The Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. The Code Case has been incorporated into Section III of the Code as Division 3, Subsection WC, Class SC Storage Containments, and will be published in the 2005 Addenda. This paper provides an informative background and insight for these rules to provide Owners, regulators, designers, and fabricators with a more comprehensive understanding of the technical basis for these rules. (authors)

Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

Energy Technology Data Exchange (ETDEWEB)

Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

Comparative evaluation of coating techniques for the corrosion protection of disposal container for spent nuclear fuel

International Nuclear Information System (INIS)

Chun, Kwan Sik; Kim, Sung Soo; Park, Chong Mook; Choi, Jong Won

2005-02-01

To propose a suitable coating technique to prevent corrosion on metal or metal alloys of a waste container to be used for the disposal of spent nuclear fuel, several methods related to spray coating and vapor deposition techniques have been comparatively evaluated, based on some major factors recommended. From these comparative results, it can be suggested that the best coating methods among the existing techniques in Korea would be HVOF and low pressure plasma spray. Even though the surface of the container coated by these methods would be coated, pores could be remained in the coated film. And therefore post-treatment methods for eliminating the pores have been briefly introduced to keep the life time of the container. The other techniques, the cold spray and hollow cathode discharge, may become excellent coating methods in the future if they are extensively researched to apply for coating on the container. An optimal process among the recommended methods should be selected by considering the state of container, such as an empty or a loaded container, and also related coating materials. For the support to this, the characteristics of the coating materials and the coated films and the durability of this film under a repository condition should be analyzed in detail

Low contaminant formic acid fuel for direct liquid fuel cell

Science.gov (United States)

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

2009-11-17

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

Microbially influenced corrosion of copper nuclear fuel waste containers in a Canadian disposal vault

Energy Technology Data Exchange (ETDEWEB)

King, F

1996-11-01

An assessment of the potential for microbially influenced corrosion (MIC) of copper nuclear fuel waste containers in a Canadian disposal vault is presented. The assessment is based on a consideration of the microbial activity within a disposal vault, the reported cases of MIC of Cu alloys in the literature and the known corrosion behaviour of Cu. Because of the critical role of biofilms in the reported cases of MIC, their formation and properties are discussed in detail. Next, the literature on the MIC of Cu alloys is briefly reviewed. The various MIC mechanisms proposed are critically discussed and the implications for the corrosion of Cu containers considered. In the majority of literature cases, MIC depends on alternating aerated and deaerated environments, with accelerated corrosion being observed when fresh aerated water replaces stagnant water, e.g., the MIC of Cu-Ni heat exchangers in polluted seawater and the microbially influenced pitting of Cu water pipes. Finally, because of the predominance of corrosion by sulphate-reducing bacteria (SRB) in the MIC literature, the abiotic behaviour of Cu alloys in sulphide solutions is also reviewed. The effect of the evolving environment in a disposal vault on the extent and location of microbial activity is discussed. Biofilm formation on the container surface is considered unlikely throughout the container lifetime, but especially initially when the environmental conditions will be particularly aggressive. Microbial activity in areas of the vault away from the container is possible, however. Corrosion of the container could then occur if microbial metabolic by-products diffuse to the container surface. Sulphide, produced by the action of SRB are considered to be the most likely cause of container corrosion. It is concluded that the only likely form of MIC of Cu containers will result from sulphide produced by SRB diffusing to the container surface. A modelling procedure for predicting the extent of corrosion is

Microbially influenced corrosion of copper nuclear fuel waste containers in a Canadian disposal vault

International Nuclear Information System (INIS)

King, F.

1996-11-01

An assessment of the potential for microbially influenced corrosion (MIC) of copper nuclear fuel waste containers in a Canadian disposal vault is presented. The assessment is based on a consideration of the microbial activity within a disposal vault, the reported cases of MIC of Cu alloys in the literature and the known corrosion behaviour of Cu. Because of the critical role of biofilms in the reported cases of MIC, their formation and properties are discussed in detail. Next, the literature on the MIC of Cu alloys is briefly reviewed. The various MIC mechanisms proposed are critically discussed and the implications for the corrosion of Cu containers considered. In the majority of literature cases, MIC depends on alternating aerated and deaerated environments, with accelerated corrosion being observed when fresh aerated water replaces stagnant water, e.g., the MIC of Cu-Ni heat exchangers in polluted seawater and the microbially influenced pitting of Cu water pipes. Finally, because of the predominance of corrosion by sulphate-reducing bacteria (SRB) in the MIC literature, the abiotic behaviour of Cu alloys in sulphide solutions is also reviewed. The effect of the evolving environment in a disposal vault on the extent and location of microbial activity is discussed. Biofilm formation on the container surface is considered unlikely throughout the container lifetime, but especially initially when the environmental conditions will be particularly aggressive. Microbial activity in areas of the vault away from the container is possible, however. Corrosion of the container could then occur if microbial metabolic by-products diffuse to the container surface. Sulphide, produced by the action of SRB are considered to be the most likely cause of container corrosion. It is concluded that the only likely form of MIC of Cu containers will result from sulphide produced by SRB diffusing to the container surface. A modelling procedure for predicting the extent of corrosion is

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

Process for introducing radioactive articles into a transport and/or storage container and transporting and/or storing the container and later extraction of the article from the container, and container for transporting and/or storing radioactive articles

International Nuclear Information System (INIS)

Vox, A.J.

1979-01-01

The articles, for example fuel elements, are introduced into the container and the remaining space inside the container is filled with lead, a salt or a mixture of salts of eutectic composition, which freezes at ambient temperature. This makes dry transport possible. To extract the fuel elements, it is sufficient to heat the container, which softens the protective and shielding material. The salt or mixture of salts is suitable for thermal conduction. (DG) [de

Nuclear fuel assemblies and fuel pins usable in such assemblies

International Nuclear Information System (INIS)

Jolly, R.

1982-01-01

A novel end cap for a nuclear fuel assembly is described in detail. It consists of a trisection arrangement which is received within a cell of a cellular grid. The cell contains abutment means with which the trisection comes into abutment. The grid also contains an abutment means for preventing the trisections from being inserted into the cell in an incorrect orientation. The present design allows fuel pins to be securely held in a hold-down grid of a sub-assembly. The design also allows easier dis-assembly of the swollen and embrittled fuel pins prior to reprocessing. (U.K.)

Method of storing the fuel storage pot in a fuel storage tank for away-from-reactor-storage

International Nuclear Information System (INIS)

Ishiguro, Jun-ichi.

1980-01-01

Purpose: To prevent the contact of sodium in the away-from-reactor-storage fuel storage tank with sodium in a fuel storage pool having radioactivity ana always retain clean state therein. Method: Sodium is filled in a container body of the away-from-reactor-storage fuel storage tank, and a conduit, a cycling pump, and cooling means are disposed to form a sodium coolant cycling loop. The fuel storage pool is so stored in the container body that the heat of the pool is projected from the liquid surface of the sodium in the container. Therefore, the sodium in the container is isolated from the sodium in the pool containing strong radioactivity to prevent contact of the former sodium from the latter sodium. (Sekiya, K.)

Assessment of the feasibility of indefinite containment of canadian nuclear fuel wastes; Evaluation de la faisabilite du confinement illimite des dechets de combustible nucleaire canadiens

Energy Technology Data Exchange (ETDEWEB)

Shoesmith, D.W.; King, F.; Ikeda, B.M.

1995-05-01

This report presents an analysis of the expected corrosion behavior of nuclear fuel waste containers in a conceptual Canadian disposal vault. The container materials considered are dilute Ti alloys (Grades-2, -12 and -16) and oxygen-free copper.

Crushing method for nuclear fuel powder

International Nuclear Information System (INIS)

Hasegawa, Shin-ichi; Tsuchiya, Haruo.

1997-01-01

A crushing medium is contained in mill pots disposed at the circumferential periphery of a main axis. The diameter of each mill pot is determined such that powdery nuclear fuels containing aggregated powders and ground and mixed powders do not reach criticality. A plurality of mill pots are revolved in the direction of the main axis while each pots rotating on its axis. Powdery nuclear fuels containing aggregated powders are conveyed to a supply portion of the moll pot, and an inert gas is supplied to the supply portion. The powdery nuclear fuels are supplied from the supply portion to the inside of the mill pots, and the powdery nuclear fuels containing aggregated powders are crushed by centrifugal force caused by the rotation and the revolving of the mill pots by means of the crushing medium. UO 2 powder in uranium oxide fuels can be crushed continuously. PuO 2 powder and UO 2 powder in MOX fuels can be crushed and mixed continuously. (I.N.)

Nuclear fuel element

International Nuclear Information System (INIS)

Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

1980-01-01

Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

Nuclear-fuel-cycle education: Module 5. In-core fuel management

International Nuclear Information System (INIS)

Levine, S.H.

1980-07-01

The purpose of this project was to develop a series of educational modules for use in nuclear-fuel-cycle education. These modules are designed for use in a traditional classroom setting by lectures or in a self-paced, personalized system of instruction. This module on in-core fuel management contains information on computational methods and theory; in-core fuel management using the Virginia Polytechnic Institute and State University computer modules; pressurized water reactor in-core fuel management; boiling water reactor in-core fuel management; and in-core fuel management for gas-cooled and fast reactors

Fuel transfer system for a nuclear reactor

International Nuclear Information System (INIS)

Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

1977-01-01

Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

Systems and processes for conversion of ethylene feedstocks to hydrocarbon fuels

Science.gov (United States)

Lilga, Michael A.; Hallen, Richard T.; Albrecht, Karl O.; Cooper, Alan R.; Frye, John G.; Ramasamy, Karthikeyan Kallupalayam

2018-04-03

Systems, processes, and catalysts are disclosed for obtaining fuel and fuel blends containing selected ratios of open-chain and closed-chain fuel-range hydrocarbons suitable for production of alternate fuels including gasolines, jet fuels, and diesel fuels. Fuel-range hydrocarbons may be derived from ethylene-containing feedstocks and ethanol-containing feedstocks.

Systems and processes for conversion of ethylene feedstocks to hydrocarbon fuels

Science.gov (United States)

Lilga, Michael A.; Hallen, Richard T.; Albrecht, Karl O.; Cooper, Alan R.; Frye, John G.; Ramasamy, Karthikeyan Kallupalayam

2017-09-26

Systems, processes, and catalysts are disclosed for obtaining fuels and fuel blends containing selected ratios of open-chain and closed-chain fuel-range hydrocarbons suitable for production of alternate fuels including gasolines, jet fuels, and diesel fuels. Fuel-range hydrocarbons may be derived from ethylene-containing feedstocks and ethanol-containing feedstocks.

Storage device of reactor fuel

International Nuclear Information System (INIS)

Nakamura, Masaaki.

1997-01-01

The present invention concerns storage of spent fuels and provides a storage device capable of securing container-cells in shielding water by remote handling and moving and securing the container-cells easily. Namely, a horizontal support plate has a plurality of openings formed in a lattice like form and is disposed in a pit filled with water. The container-cell has a rectangular cross section, and is inserted and disposed vertically in the openings. Securing members are put between the container-cells above the horizontal support plate, and constituted so as to be expandable from above by remote handling. The securing member is preferably comprised of a vertical screw member and an expandable urging member. Since securing members for securing the container-cells for incorporating reactor fuels are disposed to the horizontal support plate controllable from above by the remote handling, fuel storage device can be disposed without entering into a radiation atmosphere. The container-cells can be settled and exchanged easily after starting of the use of a fuel pit. (I.S.)

Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

International Nuclear Information System (INIS)

Durazzo, M.

1985-01-01

The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

Alteration in fuel processing at Tokai Works of Mitsubishi Nuclear Fuel Co., Ltd

International Nuclear Information System (INIS)

1977-01-01

The report of the Committee on Examination of Nuclear Fuel Safety to the Atomic Energy Commission of Japan concerning the alteration is given, which is attached to the reply from the commission to the prime minister, and its safety was confirmed. The alterations are installation of the storage for transport containers containing fuel assemblies, construction of radiation control and other buildings; and improvement and installation of the facilities for chemical-processing, pellet fabrication, fuel assembling, and storage. (Mori, K.)

Nuclear fuel preheating system

International Nuclear Information System (INIS)

Andrea, C.

1975-01-01

A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

Iron-containing N-doped carbon electrocatalysts for the cogeneration of hydroxylamine and electricity in a H-2-NO fuel cell

NARCIS (Netherlands)

Daems, Nick; Sheng, Xia; Alvarez-Gallego, Yolanda; Vankelecom, Ivo F. J.; Pescarmona, Paolo P.

2016-01-01

Iron-containing N-doped carbon materials were investigated as electrocatalysts for the cogeneration of hydroxylamine (NH2OH) and electricity in a H-2-NO fuel cell. This electrochemical route for the production of hydroxylamine is a greener alternative to the present industrial synthesis, because it

Crosslinked polybenzimidazoles containing branching structure as membrane materials with excellent cell performance and durability for fuel cell applications

Science.gov (United States)

Hu, Meishao; Ni, Jiangpeng; Zhang, Boping; Neelakandan, Sivasubramaniyan; Wang, Lei

2018-06-01

Crosslinking is an effective method to improve the properties of high temperature proton exchange membranes based on polybenzimidazole. However, the compact structure of crosslinked polybenzimidazole hinders the phosphoric acid absorption of the membranes, resulting in a relatively poor fuel cell performance. Recently, we find that branched polymers can absorb more phosphoric acid with a larger free volume, but suffer from deteriorated mechanical strength. In this work, a new method is proposed to obtain excellent over-all properties of high temperature proton exchange membranes. A series of crosslinked polybenzimidazoles containing branching structure as membrane materials are successfully prepared for the first time. Compared with conventional crosslinked membranes, these crosslinked polybenzimidazole membranes containing branching structure exhibit a higher phosphoric acid doping level and proton conductivity, improved durability, lower swelling rate and comparable mechanical strength. In particular, the fuel cell base on the crosslinked and branched membrane with a 10% ratio of crosslinker in non-humidified hydrogen/air at 160 °C achieves a power density of 404 mW cm-2. The results indicate that the combination of crosslinking and branching is an effective approach to improve the properties of polybenzimidazole membrane materials.

Fuel Cell Vehicle Basics | NREL

Science.gov (United States)

Fuel Cell Vehicle Basics Fuel Cell Vehicle Basics Researchers are developing fuel cells that can be silver four-door sedan being driven on a roadway and containing the words "hydrogen fuel cell electric" across the front and rear doors. This prototype hydrogen fuel cell electric vehicle was

Carbon fuel particles used in direct carbon conversion fuel cells

Science.gov (United States)

Cooper, John F.; Cherepy, Nerine

2012-10-09

A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

Carbon Fuel Particles Used in Direct Carbon Conversion Fuel Cells

Science.gov (United States)

Cooper, John F.; Cherepy, Nerine

2008-10-21

A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

Gelled fuel simulant

International Nuclear Information System (INIS)

Christy, J.; Hiser, E.J.; Sippel, N.J.

1980-01-01

A relatively stable inert simulant formulation for a hazardous metallized fuel has the density, shear rate and yield stress of the duplicated fuel. This formulation provides inexpensive and safe testing of exploratory hydraulic studies, or testing of the mechanical strength of containers, plumbing, etc., in which the metallized fuels are to be used

Concept for premixed combustion of hydrogen-containing fuels in gas turbines; Konzept zur vorgemischten Verbrennung wasserstoffhaltiger Brennstoffe in Gasturbinen

Energy Technology Data Exchange (ETDEWEB)

Mayer, Christoph

2012-07-19

One of the main challenges for future gas turbines and their combustion systems is to provide fuel flexibility. The fuel range is expected to reach from the lowly reactive natural gas to highly reactive hydrogen-containing syngases. The objective of the project in which this work was pursued is to develop such a combustion system. The burner has to ensure premixed operation with an aerodynamically stabilized flame. The focus of this work is on characterizing and optimizing the operational safety of the system, but also on ensuring sufficientmixing and lowemissions. A burner and fuel injection design is achieved that leads not only to emissions far below the permissible values, but also to flashback safety for hydrogen combustion that comes close to the theoretically achievable maximum at atmospheric pressure conditions. In this design flashback due to combustion-induced vortex breakdown and wall boundary layer flashback is avoided. Flashback only takes place when the flow velocity reaches the flame velocity.

Formulation, Casting, and Evaluation of Paraffin-Based Solid Fuels Containing Energetic and Novel Additives for Hybrid Rockets

Science.gov (United States)

Larson, Daniel B.; Desain, John D.; Boyer, Eric; Wachs, Trevor; Kuo, Kenneth K.; Borduin, Russell; Koo, Joseph H.; Brady, Brian B.; Curtiss, Thomas J.; Story, George

2012-01-01

This investigation studied the inclusion of various additives to paraffin wax for use in a hybrid rocket motor. Some of the paraffin-based fuels were doped with various percentages of LiAlH4 (up to 10%). Addition of LiAlH4 at 10% was found to increase regression rates between 7 - 10% over baseline paraffin through tests in a gaseous oxygen hybrid rocket motor. Mass burn rates for paraffin grains with 10% LiAlH4 were also higher than those of the baseline paraffin. RDX was also cast into a paraffin sample via a novel casting process which involved dissolving RDX into dimethylformamide (DMF) solvent and then drawing a vacuum on the mixture of paraffin and RDX/DMF in order to evaporate out the DMF. It was found that although all DMF was removed, the process was not conducive to generating small RDX particles. The slow boiling generated an inhomogeneous mixture of paraffin and RDX. It is likely that superheating the DMF to cause rapid boiling would likely reduce RDX particle sizes. In addition to paraffin/LiAlH4 grains, multi-walled carbon nanotubes (MWNT) were cast in paraffin for testing in a hybrid rocket motor, and assorted samples containing a range of MWNT percentages in paraffin were imaged using SEM. The fuel samples showed good distribution of MWNT in the paraffin matrix, but the MWNT were often agglomerated, indicating that a change to the sonication and mixing processes were required to achieve better uniformity and debundled MWNT. Fuel grains with MWNT fuel grains had slightly lower regression rate, likely due to the increased thermal conductivity to the fuel subsurface, reducing the burning surface temperature.

BWR fuel performance

International Nuclear Information System (INIS)

Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

1979-01-01

The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

On the origin of metal impurities in content of lava-like fuel-containing materials of Chornobyl NSC-Shelter object

Directory of Open Access Journals (Sweden)

O. V. Mikhailov

2017-11-01

Full Text Available Version of the origin of material metallic impurities in silicate matrix of lava-like fuel-containing masses (LFCM, which were formed during the Chernobyl Unit 4 accident, is presented. Based on comparative quantitative characteristics of observable mass ratios of iron, chromium and nickel in different LFCM clusters and potential sources of their appearance - metal structures, the degree of impact of various factors on the formation of metal components in Chernobyl corium (MCC was given. It was concluded that initial MCC composition was formed on the basis of 08X18H10T stainless steel melt, from which the elements of fuel channel structure and lower water pipelines were manufactured.

The CANDU 9 fuel transfer system

International Nuclear Information System (INIS)

Keszthelyi, Z.G.; Morikawa, D.T.

1996-01-01

The CANDU 9 fuel transfer system is based on the CANDU 6 and the Ontario Hydro Darlington NGD designs, modified to suit the CANDU 9 requirements. The CANDU 9 new fuel transfer system is very similar to the CANDU 6, with modifications to allow new fuel loading from outside containment, similar to Darlington. The CANDU 9 irradiated fuel transfer system is based on the Darlington irradiated fuel transfer system, with modifications to meet the more stringent containment requirements, improve performance, and match station layout. (author). 2 refs., 6 figs

The CANDU 9 fuel transfer system

Energy Technology Data Exchange (ETDEWEB)

Keszthelyi, Z G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada); Morikawa, D T [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1997-12-31

The CANDU 9 fuel transfer system is based on the CANDU 6 and the Ontario Hydro Darlington NGD designs, modified to suit the CANDU 9 requirements. The CANDU 9 new fuel transfer system is very similar to the CANDU 6, with modifications to allow new fuel loading from outside containment, similar to Darlington. The CANDU 9 irradiated fuel transfer system is based on the Darlington irradiated fuel transfer system, with modifications to meet the more stringent containment requirements, improve performance, and match station layout. (author). 2 refs., 6 figs.

Treat upgrade fuel fabrication

International Nuclear Information System (INIS)

Davidson, K.V.; Schell, D.H.

1979-01-01

An extrusion and thermal treatment process was developed to produce graphite fuel rods containing a dispersion of enriched UO 2 . These rods will be used in an upgraded version of the Transient Reactor Test Facility (TREAT). The improved fuel provides a higher graphite matrix density, better fuel dispersion and higher thermal capabilities than the existing fuel

DUPIC fuel fabrication using spent PWR fuels at KAERI

Energy Technology Data Exchange (ETDEWEB)

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

2000-12-01

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

The future supply of and demand for candidate materials for the fabrication of nuclear fuel waste disposal containers

International Nuclear Information System (INIS)

Grover, L.K.

1990-01-01

This report summarizes the findings of a literature survey carried out to assess the future world supply of and demand for titanium, copper and lead. These metals are candidate materials for the fabrication of containers for the immobilization and disposal of Canada's nuclear used-fuel waste for a reference Used-fuel Disposal Centre. Such a facility may begin operation by approximately 2020, and continue for about 40 years. The survey shows that the world has abundant supplies of titanium minerals (mostly in the form of ilmenite), which are expected to last up to at least 2110. However, for copper and lead the balance between supply and demand may warrant increased monitoring beyond the year 2000. A number of factors that can influence future supply and demand are discussed in the report

Fabrication of internally instrumented reactor fuel rods

International Nuclear Information System (INIS)

Schmutz, J.D.; Meservey, R.H.

1975-01-01

Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

Method for compacting spent nuclear reactor fuel rods

International Nuclear Information System (INIS)

Wachter, W.J.

1988-01-01

In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

Stability of MOF-5 in a hydrogen gas environment containing fueling station impurities

DEFF Research Database (Denmark)

Ming, Yang; Purewal, Justin; Yang, Jun

2016-01-01

in the hydrogen fuel stream. Hydrogen intended for use in fuel cell vehicles should satisfy purity standards, such as those outlined in SAE J2719. This standard limits the concentration of certain species in the fuel stream based primarily on their deleterious effects on PEM fuel cells. However, the impact...

Transport of irradiated nuclear fuel

International Nuclear Information System (INIS)

1980-01-01

In response to public interest in the transport by rail through London of containers of irradiated fuel elements on their way from nuclear power stations to Windscale, the Central Electricity Generating Board and British Rail held three information meetings in London in January 1980. One meeting was for representatives of London Borough Councils and Members of Parliament with a known interest in the subject, and the others were for press, radio and television journalists. This booklet contains the main points made by the principal speakers from the CEGB and BR. (The points covered include: brief description of the fuel cycle; effect of the fission process in producing plutonium and fission products in the fuel element; fuel transport; the fuel flasks; protection against accidents; experience of transporting fuel). (U.K.)

1990 fuel cell seminar: Program and abstracts

Energy Technology Data Exchange (ETDEWEB)

1990-12-31

This volume contains author prepared short resumes of the presentations at the 1990 Fuel Cell Seminar held November 25-28, 1990 in Phoenix, Arizona. Contained herein are 134 short descriptions organized into topic areas entitled An Environmental Overview, Transportation Applications, Technology Advancements for Molten Carbonate Fuel Cells, Technology Advancements for Solid Fuel Cells, Component Technologies and Systems Analysis, Stationary Power Applications, Marine and Space Applications, Technology Advancements for Acid Type Fuel Cells, and Technology Advancement for Solid Oxide Fuel Cells.

Fuel transfer machine

International Nuclear Information System (INIS)

Bernstein, I.

1978-01-01

A nuclear fuel transfer machine for transferring fuel assemblies through the fuel transfer tube of a nuclear power generating plant containment structure is described. A conventional reversible drive cable is attached to the fuel transfer carriage to drive it horizontally through the tube. A shuttle carrying a sheave at each end is arranged in parallel with the carriage to also travel into the tube. The cable cooperating with the sheaves permit driving a relatively short fuel transfer carriage a large distance without manually installing sheaves or drive apparatus in the tunnel. 8 claims, 3 figures

Spent nuclear fuel storage vessel

International Nuclear Information System (INIS)

Watanabe, Yoshio; Kashiwagi, Eisuke; Sekikawa, Tsutomu.

1997-01-01

Containing tubes for containing spent nuclear fuels are arranged vertically in a chamber. Heat releasing fins are disposed horizontal to the outer circumference of the containing tubes for rectifying cooling air and promoting cooling of the containing tubes. Louvers and evaporation sides of heat pipes are disposed at a predetermined distance in the chamber. Cooling air flows from an air introduction port to the inside of the chamber and takes heat from the containing tubes incorporated with heat generating spent nuclear fuels, rising its temperature and flows off to an air exhaustion exit. The direction for the rectification plate of the louver is downward from a horizontal position while facing to the air exhaustion port. Since the evaporation sides of the heat pipes are disposed in the inside of the chamber and the condensation side of the heat pipes is disposed to the outside of the chamber, the thermal energy can be recovered from the containing tubes incorporated with spent nuclear fuels and utilized. (I.N.)

Operation and maintenance of spent fuel storage and transportation casks/containers

International Nuclear Information System (INIS)

2007-01-01

Member States have a growing need for casks for spent fuel storage and transportation. A variety of casks has been developed and is in use at an increasing number of sites. This has resulted in an accumulation of experience that will provide valuable information for other projects in spent fuel management. This publication provides a comprehensive review of information on the cask operation and maintenance associated with spent fuel storage. It draws upon generic knowledge from industrial experience and applications and is intended to serve as a basis for better planning and implementation in future projects

Solar Fuel Generator

Science.gov (United States)

Lewis, Nathan S. (Inventor); West, William C. (Inventor)

2017-01-01

The disclosure provides conductive membranes for water splitting and solar fuel generation. The membranes comprise an embedded semiconductive/photoactive material and an oxygen or hydrogen evolution catalyst. Also provided are chassis and cassettes containing the membranes for use in fuel generation.

40 CFR 79.56 - Fuel and fuel additive grouping system.

Science.gov (United States)

2010-07-01

... industry-sponsored or other independent brokering arrangements. (3) Manufacturers who enroll a fuel or fuel... Specification for Automotive Spark-Ignition Engine Fuel”, used to define the general characteristics of gasoline... shall be chemical-grade quality, at a minimum, and shall not contain a significant amount of other...

Synthesis of fuels and feedstocks

Energy Technology Data Exchange (ETDEWEB)

Sutton, Andrew D.; Brooks, Ty; Jenkins, Rhodri; Moore, Cameron; Staples, Orion

2017-10-10

Disclosed herein are embodiments of a method for making fuels and feedstocks from readily available alcohol starting materials. In some embodiments, the method concerns converting alcohols to carbonyl-containing compounds and then condensing such carbonyl-containing compounds together to form oligomerized species. These oligomerized species can then be reduced using by-products from the conversion of the alcohol. In some embodiments, the method further comprises converting saturated, oligomerized, carbonyl-containing compounds to aliphatic fuels.

Stabilizing motor fuels

Energy Technology Data Exchange (ETDEWEB)

1935-07-12

Motor fuel is stabilized by adding less than 2% of a tar fraction from peat, coal, torbanite or shale, said fraction containing sufficient constituents boiling between 200 and 325/sup 0/C, to inhibit gum formation. Low-temperature coal-tar fractions are specified. The preferred boiling ranges are from 225 or 250/sup 0/ to 275/sup 0/C. In examples, the quantity added was 0.01%. The fuel may be a cracked distillate of gasoline boiling-point range or containing gasoline, and may contain relatively large proportions of di- and tri-olefines. The material added to the fuel may be (1) the tar fraction itself; (2) its alkali-soluble constituents; (3) its acid-soluble constituents; (4) a mixture of (2) and (3); (5) a blend of (2), (3) or (4) with a normal tar fraction; (6) the residue after extraction with alkali; (7) the residue after extraction with acid and alkali.

Progress in Chemical Kinetic Modeling for Surrogate Fuels

Energy Technology Data Exchange (ETDEWEB)

Pitz, W J; Westbrook, C K; Herbinet, O; Silke, E J

2008-06-06

Gasoline, diesel, and other alternative transportation fuels contain hundreds to thousands of compounds. It is currently not possible to represent all these compounds in detailed chemical kinetic models. Instead, these fuels are represented by surrogate fuel models which contain a limited number of representative compounds. We have been extending the list of compounds for detailed chemical models that are available for use in fuel surrogate models. Detailed models for components with larger and more complicated fuel molecular structures are now available. These advancements are allowing a more accurate representation of practical and alternative fuels. We have developed detailed chemical kinetic models for fuels with higher molecular weight fuel molecules such as n-hexadecane (C16). Also, we can consider more complicated fuel molecular structures like cyclic alkanes and aromatics that are found in practical fuels. For alternative fuels, the capability to model large biodiesel fuels that have ester structures is becoming available. These newly addressed cyclic and ester structures in fuels profoundly affect the reaction rate of the fuel predicted by the model. Finally, these surrogate fuel models contain large numbers of species and reactions and must be reduced for use in multi-dimensional models for spark-ignition, HCCI and diesel engines.

Third international conference on CANDU fuel

International Nuclear Information System (INIS)

Boczar, Peter

1992-01-01

These proceedings contain full texts of all 49 papers from the ten sessions and the banquet address. The sessions were on the following subjects: International experience and programs; Fuel behaviour and operating experience; Fuel modelling; Fuel design; Advanced fuel and fuel cycle technology; AECL's concept for the disposal of nuclear fuel waste. The individual papers have been abstracted separately

Analysis of radiation doses from operation of postulated commercial spent fuel transportation systems: Analysis of a system containing a monitored retrievable storage facility

International Nuclear Information System (INIS)

Smith, R.I.; Daling, P.M.; Faletti, D.W.

1992-04-01

This addendum report extends the original study of the estimated radiation doses to the public and to workers resulting from transporting spent nuclear fuel from commercial nuclear power reactor stations through the federal waste management system (FWMS), to a system that contains a monitored retrievable storage (MRS) facility. The system concepts and designs utilized herein are consistent with those used in the original study (circa 1985--1987). Because the FWMS design is still evolving, the results of these analyses may no longer apply to the design for casks and cask handling systems that are currently being considered. Four system scenarios are examined and compared with the reference No-MRS scenario (all spent fuel transported directly from the reactors to the western repository in standard-capacity truck and rail casks). In Scenarios 1 and 2, an MRS facility is located in eastern United States and ships either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters. In Scenarios 3 and 4, an MRS facility is located in the western United States and ship either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters

Spent nuclear fuel transport problems

International Nuclear Information System (INIS)

Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.I.

1977-01-01

The paper considers the problems of shipping spent fuel from nuclear power stations to reprocessing plants and also the principal ways of solving these problems with a view to achieving maximum economy and safety in transport. The increase in the number of nuclear power plants in the USSR will entail an intensification of spent-fuel shipments. Higher burnup and the need to reduce cooling time call for heavier and more complex shipping containers. The problem of shipping spent fuel should be tackled comprehensively, bearing in mind the requirements of safety and economy. One solution to these problems is to develop rational and cheap designs of such containers. In addition, the world-wide trend towards more thorough protection of the environment against pollution and of the health of the population requires the devotion of constant attention to improving the reliability and safety of shipments. The paper considers the prospects for nuclear power development in the USSR and in other member countries of the CMEA (1976-1980), the composition and design of some Soviet packaging assemblies, the appropriate cooling time for spent fuel from thermal reactor power stations, procedures for reducing fuel-shipping costs, some methodological problems of container calculation and design, and finally problems of testing and checking containers on test rigs. (author)

Dual winch nuclear fuel transfer system providing more reliable fuel transfer during refueling operations

International Nuclear Information System (INIS)

Meuschke, R.E.; Harper, M.J.; Stefko, D.J.

1991-01-01

This paper describes a nuclear power plant having an auxiliary building, a containment building having the wall, a track extending through a transfer tube within the containment wall, and a fuel transfer system for moving fuel assemblies along the track between the auxiliary building side and the containment building side of the containment wall. It comprises: a car having wheels for movement along spaced rails of the track and further having a carrying basket for one or more fuel assemblies; winch means located on the auxiliary building side of the containment wall and above the water level existing over the track during refueling operations to drive the car along the track; first cable means and second cable means extending substantially vertically downward from the winch means to the tack level; first sheave means for directing the first and the second cable means substantially in the horizontal direction along the track; means for securing the first cable means to the car so that winch pulling force on the first cable means drives the car away from the containment building; second sheave means located near the containment end of the transfer tube

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi; Masumi, Ryoji; Soneda, Hideo.

1994-01-01

A fuel assembly comprises a plurality of fuel rods filled with nuclear fuels, a plurality of burnable poison rods incorporated with burnable poisons, and water rods which can vary the height in the tube depending on the coolant flow rate flown into the assembly. The amount of entire burnable poisons of the burnable poison-containing rods in adjacent with the water rods is smaller than the amount of entire burnable poisons in the burnable poison containing rods not in adjacent with the water rods. Then the average concentration of burnable poisons in the axial upper half region is made smaller than the average concentration of the burnable poisons at the axial lower half region. Further, a burnable poison concentration at the upper half region of at least one of burnable poison-containing rods in adjacent with the water rods is made lower than the burnable poison concentration in the lower half region. Since this can fasten the combustion of the burnable poisons, a fuel assembly having good fuel economy can be attained. (I.N.)

Fuel rod with axial regions of annular and standard fuel pellets

International Nuclear Information System (INIS)

Freeman, T.R.

1991-01-01

This patent describes a fuel rod for use in a nuclear reactor fuel assembly. It comprises: an elongated hollow cladding tube; a pair of end plugs connected to and sealing the cladding tube at opposite ends of thereof; and an axial stack of fuel pellets contained in and extending between the end plugs at the opposite ends of the tube, all of the fuel pellets contained in the tube being composed of fissile material being enriched above the level of natural enrichment; the fuel pellets in the stack thereof being provided in an arrangement of axial regions. The arrangement of axial regions including a pair of first axial regions defined respectively at the opposite ends of the pellet stack adjacent to the respective end plugs. The pellets in the first axial regions being identical in number and having annular configurations with an annulus of a first void size. The arrangement of axial regions also including another axial region defined between the first axial regions, some of the pellets in the another axial region having solid configurations

Passive containment system

International Nuclear Information System (INIS)

Kleimola, F.W.

1977-01-01

Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

Fuel Chemistry Research | Transportation Research | NREL

Science.gov (United States)

Fuel Chemistry Research Fuel Chemistry Research Photo of a hand holding a beaker containing a clear oils. Photo by Dennis Schroeder, NREL NREL's fuel chemistry research explores how biofuels, advanced , emissions control catalysts, and infrastructure materials. Results from NREL's fuel chemistry studies feed

Containment integrity analysis with SAMPSON/DCRA module

International Nuclear Information System (INIS)

Hosoda, Seigo; Shirakawa, Noriyuki; Naitoh, Masanori

2006-01-01

The integrity of PWR containment under a severe accident is analyzed using the debris concrete reaction analysis code. If core fuels melt through the pressure vessel and the debris accumulates on the reactor cavity of a lower part of containment, its temperature continues to rise due to decay heat and the debris ablates the concrete floor. In case that cooling water is issued into the containment cavity and the amount of debris is limited to 30% of core fuels, our analyses showed that the debris could be cooled and frozen so that integrity of containment could hold. (author)

Segmented fuel and moderator rod

International Nuclear Information System (INIS)

Doshi, P.K.

1987-01-01

This patent describes a continuous segmented fuel and moderator rod for use with a water cooled and moderated nuclear fuel assembly. The rod comprises: a lower fuel region containing a column of nuclear fuel; a moderator region, disposed axially above the fuel region. The moderator region has means for admitting and passing the water moderator therethrough for moderating an upper portion of the nuclear fuel assembly. The moderator region is separated from the fuel region by a water tight separator

Orimulsion containment and recovery

International Nuclear Information System (INIS)

Sommerville, M.

1999-01-01

This paper focuses on the need for examination of Orimulsion fuel and its spill behaviour in the light of the anticipated increase in consumption of this fuel which comprises bitumen dispersed in water with addition of a small amount of surfactant. The behaviour and fate of Orimulsion at sea, and observations from experimental and sea trials are examined. The identification of spill control techniques, spill detection, the predictive modeling of the spill and response, sub-surface plume measurement, and containment and deflection are considered. Recovery of the bitumen produced from an Orimulsion spill, combined containment and recovery, dispersed Orimulsion, and beach cleaning are addressed. The properties of Orimulsion are tabulated. (UK)

Nuclear fuel assembly

International Nuclear Information System (INIS)

Wakamatsu, Mitsuo.

1974-01-01

Object: To improve a circulating flow passage of coolant so as to be able to accurately detect the temperature of coolant, rare gases contained, and the like. Structure: A fuel assembly comprising a flow regulating lattice provided with a plurality of communication holes in an axial direction, said lattice being positioned at the upper end of an outer tube in which nuclear fuel elements are received, and a neutron shielding body having a plurality of spiral coolant flow passages disposed between the lattice and the nuclear fuel elements, whereby a coolant comprised of liquid sodium or the like, which moves up passing through the coolant flow passages and the flow regulating passage, is regulated and passed through a detector mounted at the upper part of the flow regulating lattice to detect coolant temperature, flow rate, and rare gases or the like as the origin of nuclear fission contained in the coolant due to breakage of fuel elements. (Kamimura, M.)

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

Energy Technology Data Exchange (ETDEWEB)

Barranco R, F.

2015-07-01

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

Report of Post Irradiation Examination for Dry Process Fuel

International Nuclear Information System (INIS)

Par, Jang Jin; Jung, I. H.; Kang, K. H.; Moon, J. S.; Lee, C. R.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Yoo, B. O.; Jung, Y. H.; Choo, Y. S.

2006-08-01

The spent PWR fuel typically contains 0.9 wt.% of fissile uranium and 0.6 wt.% of fissile plutonium, which exceeds the natural uranium fissile content of 0.711 wt.%. The neutron economy of a CANDU reactor is sufficient to utilize the DUPIC fuel, even though the neutron-absorbing fission products contained in the spent PWR fuel were remained in the DUPIC fuel. The DUPIC fuel cycle offers advantages to the countries operating both the PWR and CANDU reactors, such as saving the natural uranium, reducing the spent fuel in both PWR and CANDU, and acquiring the extra energy by reuse of the PWR spent fuel. This report contains the results of post-irradiation examination of the DUPIC fuel irradiated four times at HANARO from May 2000 to August 2006 present except the first irradiation test of simulated DUPIC fuel at HANARO on August 1999

46 CFR 153.1025 - Motor fuel antiknock compounds.

Science.gov (United States)

2010-10-01

... 46 Shipping 5 2010-10-01 2010-10-01 false Motor fuel antiknock compounds. 153.1025 Section 153... Cargo Procedures § 153.1025 Motor fuel antiknock compounds. (a) No person may load or carry any other cargo in a containment system approved for motor fuel antiknock compounds containing lead alkyls except...

URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

Science.gov (United States)

Handwerk, J.H.; Noland, R.A.; Walker, D.E.

1957-09-10

In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

Procedure for drying humidity-containing bodies

International Nuclear Information System (INIS)

Johnson, C.R.

1976-01-01

The invention concerns a decontamination process for extracting impurities, in particular humidity and gases, from nuclear fuel rods before they are sealed and inserted into the reactor. The fuel rod, which has a small drilling hole, is placed in a low pressure container. The container is filled with a liquid drying agent which washes out the impurities. A dry inert gas (nitrogen, noble gases) is used for rinsing. Alcohols, ketones, methanol, acetone are named as drying agents. (UWI) [de

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

Status of LEU fuel development and conversion of NRU

International Nuclear Information System (INIS)

Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

1991-01-01

This paper reviews the status of the LEU conversion program and the progress made in the fuel development program over the last year. The results from post-irradiation examinations of prototype NRU fuel rods containing Al-U 3 Si dispersion fuel, and of mini-elements containing Al-U 3 Si 2 dispersion fuel, are presented. (orig.)

Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

Science.gov (United States)

Reichner, Philip; Dollard, Walter J.

1991-01-01

An electrochemical apparatus (10) is made having a generator section (22) containing axially elongated electrochemical cells (16), a fresh gaseous feed fuel inlet (28), a gaseous feed oxidant inlet (30), and at least one gaseous spent fuel exit channel (46), where the spent fuel exit channel (46) passes from the generator chamber (22) to combine with the fresh feed fuel inlet (28) at a mixing apparatus (50), reformable fuel mixture channel (52) passes through the length of the generator chamber (22) and connects with the mixing apparatus (50), that channel containing entry ports (54) within the generator chamber (22), where the axis of the ports is transverse to the fuel electrode surfaces (18), where a catalytic reforming material is distributed near the reformable fuel mixture entry ports (54).

DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

International Nuclear Information System (INIS)

Rest, J.; Hofman, G.L.

1997-01-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 -Al for various dispersion fuel element designs with the data

Novel synthesis of highly durable and active Pt catalyst encapsulated in nitrogen containing carbon for polymer electrolyte membrane fuel cell

Science.gov (United States)

Lee, Hyunjoon; Sung, Yung-Eun; Choi, Insoo; Lim, Taeho; Kwon, Oh Joong

2017-09-01

Novel synthesis of a Pt catalyst encapsulated in a N-containing carbon layer for use in a polymer electrolyte membrane fuel cell is described in this study. A Pt-aniline complex, formed by mixing Pt precursor and aniline monomer, was used as the source of Pt, C, and N. Heat treatment of the Pt-aniline complex with carbon black yielded 5 nm Pt nanoparticles encapsulated by a N-containing carbon layer originating from aniline carbonization. The synthesized Pt catalyst exhibited higher mass specific activity to oxygen reduction reaction than that shown by conventional Pt/C catalyst because pyridinic N with graphitic carbon in the carbon layer provided active sites for oxygen reduction reaction in addition to those provided by Pt. In single cell testing, initial performance of the synthesized catalyst was limited because the thick catalyst layer increased resistance related to mass transfer. However, it was observed that the carbon layer successfully prevented Pt nanoparticles from growing via agglomeration and Ostwald ripening under fuel cell operation, thereby improving durability. Furthermore, a mass specific performance of the synthesized catalyst higher than that of a conventional Pt/C catalyst was achieved by modifying the synthesized catalyst's layer thickness.

A concept to combine DOE waste minimization goals with commercial utility needs for a universal container system for spent nuclear fuel storage, transportation, and disposal

International Nuclear Information System (INIS)

Falci, F.P.; Smith, M.L.; Sorenson, K.B.

1993-01-01

The concept of storing, transporting, and disposing of spent fuel using a single package has obvious advantages. Coupling this concept with using contaminated scrap metal from the EM Complex will help reduce a significant portion of waste that would otherwise need to be packaged, stored, and disposed of as low level radioactive waste. Assuming a material of cost of $1 per pound for 800,000 tons of metal needed for universal containers, the potential material cost savings from manufacturing these containers from what would otherwise be a waste product is about $1.5 billion. Clearly, this concept is novel and has significant obstacles that need to be addressed and overcome; particularly in the regulatory arena. However, the potential benefits warrant the evaluation of the proposal on several fronts. DOE OCRWM should seriously consider the universal cask concept for management of spent fuel. DOE EM should pursue the development of melting contaminated scrap for the manufacture of casks. Finally, EM and OCRWM should cooperate on the evaluation of using EM contaminated scrap metal for the manufacture of universal casks for OCRWM spent fuel

Formulation and Testing of Paraffin-Based Solid Fuels Containing Energetic Additives for Hybrid Rockets

Science.gov (United States)

Larson, Daniel B.; Boyer, Eric; Wachs,Trevor; Kuo, Kenneth K.; Story, George

2012-01-01

Many approaches have been considered in an effort to improve the regression rate of solid fuels for hybrid rocket applications. One promising method is to use a fuel with a fast burning rate such as paraffin wax; however, additional performance increases to the fuel regression rate are necessary to make the fuel a viable candidate to replace current launch propulsion systems. The addition of energetic and/or nano-sized particles is one way to increase mass-burning rates of the solid fuels and increase the overall performance of the hybrid rocket motor.1,2 Several paraffin-based fuel grains with various energetic additives (e.g., lithium aluminum hydride (LiAlH4) have been cast in an attempt to improve regression rates. There are two major advantages to introducing LiAlH4 additive into the solid fuel matrix: 1) the increased characteristic velocity, 2) decreased dependency of Isp on oxidizer-to-fuel ratio. The testing and characterization of these solid-fuel grains have shown that continued work is necessary to eliminate unburned/unreacted fuel in downstream sections of the test apparatus.3 Changes to the fuel matrix include higher melting point wax and smaller energetic additive particles. The reduction in particle size through various methods can result in more homogeneous grain structure. The higher melting point wax can serve to reduce the melt-layer thickness, allowing the LiAlH4 particles to react closer to the burning surface, thus increasing the heat feedback rate and fuel regression rate. In addition to the formulation of LiAlH4 and paraffin wax solid-fuel grains, liquid additives of triethylaluminum and diisobutylaluminum hydride will be included in this study. Another promising fuel formulation consideration is to incorporate a small percentage of RDX as an additive to paraffin. A novel casting technique will be used by dissolving RDX in a solvent to crystallize the energetic additive. After dissolving the RDX in a solvent chosen for its compatibility

Design of double containment canister cask storage system

International Nuclear Information System (INIS)

Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

2004-01-01

Spent fuels discharged from Japanese LWR will be stored as recycled-fuel-resources in interim storage facilities. The concrete cask storage system is one of important forms for the spent fuel interim storage. In Japan, the interim storage facility will be located near the coast, therefore it is important to prevent SCC (Stress Corrosion Cracking) caused by sea salt particles and to assure the containment integrity of the canister which contains spent fuels. KEPCO, NFT and OCL have designed the double containment canister cask storage system that can assure the long-term containment integrity and monitor the containment performance without storage capacity decrease. Major features of the combined canister cask system are shown as follows: This system can survey containment integrity of dual canisters by monitoring the pressure of the gap between canisters. The primary canister has dual lids sealed by welding. The secondary canister has single lid tightened by bolts and sealed by metallic gaskets. The primary canister is contained in the transport cask during transportation, and the gap between the primary canister and the transport cask is filled with He gas. Under storage condition in the concrete cask, the primary canister is contained in the secondary canister, and the gap between these canisters is filled with helium gas. Hence this system can prevent the primary canister to contact sea salt particle in the air and from SCC. Decrease of cooling performance because of the double canister is compensated by fins fitted on the secondary canister surface. Then, this system can prevent the decrease of storage capacity determined by the fuel temperature limit. This system can assure that the primary canister will keep intact for long term storage. Therefore, in the case of pressure down of the gap between canisters, it can be considered that the secondary canister containment is damaged, and the primary canister will be transferred to another secondary canister at the

Transportation of nuclear fuel

International Nuclear Information System (INIS)

Prowse, D.R.

1979-01-01

Shipment of used fuel from nuclear reactors to a central fuel management facility is discussed with particular emphasis on the assessment of the risk to the public due to these shipments. The methods of transporting used fuel in large shipping containers is reviewed. In terms of an accident scenario, it is demonstrated that the primary risk of transport of used fuel is due to injury and death in common road accidents. The radiological nature of the used fuel cargo is, for all practical purposes, an insignificant factor in the total risk to the public. (author)

Modeling status and needs for temperature calculations within spent fuel disposal containers

International Nuclear Information System (INIS)

Sullivan, T.M.; Pescatore, C.

1989-10-01

The Brookhaven National Laboratory (BNL) Waste Materials and Environment Modeling (WMEM) Program has been assigned the task of helping the DOE formulate and certify analytical tools needed to support and/or strengthen the Waste Package Licensing strategy. One objective of the WMEM program is to perform qualitative and quantitative analyses of processes related to the internal waste package environment, e.g., temperature, radiolysis effects, presence of moisture, etc. The primary objective of this report is to present the findings of a literature review of work pertinent to predicting intact waste package internal temperatures under spent fuel isolation conditions. Therefore, it is assumed that a repository scale thermal analysis has been conducted and the exterior temperature of the waste package is known. Thus, the problem reduces to one determined by the waste package and its properties. Secondary objectives of this report are to identify key parameters and methodologies for performing the thermal analysis within intact waste containers, and identify sources of uncertainty in these calculations. 37 refs., 6 figs., 2 tabs

Fuel storage rack

International Nuclear Information System (INIS)

Mollon, L.

1977-01-01

Disclosed is a storage rack for spent nuclear fuel elements comprising a multiplicity of elongated hollow containers of uniform cross-section, preferably square,some of said containers having laterally extending continuous flanges extending between adjacent containers and defining continuous elongated chambers therebetween for the reception of neutron absorbing panels. 18 claims, 7 figures

Seal for an object containing nuclear fuel

International Nuclear Information System (INIS)

Scheuerpflug, W.; Nentwich, D.

1977-01-01

This seal which cannot be counterfeited, specially for sealing nuclear objects, e.g. fuel rods, not only makes any damage which has taken place obvious, but makes identification according to a key possible. For this purpose a minimum number of 'particles' or small bodies, which are identical but of different permeability, are fixed inside a short tube during 'loading' of the seal in a certain or an accidental sequence. The sequence of the spheres, which represents a key, can only be determined by special electromagnetic measuring equipment. On first opening the seal, this key sequence is irrevocably destroyed. (HP) [de

Elucidation of oxidation and degradation products of oxygen containing fuel components by combined use of a stable isotopic tracer and mass spectrometry.

Science.gov (United States)

Frauscher, Marcella; Besser, Charlotte; Allmaier, Günter; Dörr, Nicole

2017-11-15

In order to reveal the degradation products of oxygen-containing fuel components, in particular fatty acid methyl esters, a novel approach was developed to characterize the oxidation behaviour. Combination of artificial alteration under pressurized oxygen atmosphere, a stable isotopic tracer, and gas chromatography electron impact mass spectrometry (GC-EI-MS) was used to obtain detailed information on the formation of oxidation products of (9Z), (12Z)-octadecadienoic acid methyl ester (C18:2 ME). Thereby, biodiesel simulating model compound C18:2 ME was oxidized in a rotating pressurized vessel standardized for lubricant oxidation tests (RPVOT), i.e., artificially altered, under 16 O 2 as well as 18 O 2 atmosphere. Identification of the formed degradation products, mainly carboxylic acids of various chain lengths, alcohols, ketones, and esters, was performed by means of GC-EI-MS. Comparison of mass spectra of compounds under both atmospheres revealed not only the degree of oxidation and the origin of oxygen atoms, but also the sites of oxidative attack and bond cleavage. Hence, the developed and outlined strategy based on a gas-phase stable isotopic tracer and mass spectrometry provides insight into the degradation of oxygen-containing fuels and fuel components by means of the accurate differentiation of oxygen origin in a degradation product. Copyright © 2017 Elsevier B.V. All rights reserved.

Nuclear reactor fuel element containing an end piece for maintaining the column of fuel pellets

International Nuclear Information System (INIS)

Pajot, Jacques; Rabellino, Jacques.

1974-01-01

The nuclear reactor fuel element described has an end piece for maintaining the column of fuel pellets in position inside the element cladding. This end piece has a central compression spring one end of which presses against the pellets and the other against a plug shaped piece fitted with a seat for the spring, a conical piece with an elastic ring around it diverging towards the end in contact with the spring and a head at the opposite end. The connection between the compression spring and the pellets is through an application piece. A central bore provided in the end piece helps balance the pressure inside the element. This element is particularly intended for liquid metal cooled fast neutron reactors [fr

Fuel development and manufacturing programme in India and advanced fuel designs

International Nuclear Information System (INIS)

Das, M.; Bhardwaj, S.A.; Saxena, A.K.; Anantharaman, K.; Varma, B.P.

1995-01-01

The emphasis of self reliance in all areas of nuclear fuel cycle technology is the objective of Department of Atomic Energy, India. To achieve this aim, various organisations are working in close co-operation. This paper contains a brief summary of the work carried out in India on PHWR fuel technology

Standard guide for drying behavior of spent nuclear fuel

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

Fuel processor for fuel cell power system. [Conversion of methanol into hydrogen

Science.gov (United States)

Vanderborgh, N.E.; Springer, T.E.; Huff, J.R.

1986-01-28

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

CRITICALITY SAFETY CONTROL OF LEGACY FUEL FOUND AT 105-K WEST FUEL STORAGE BASIN

International Nuclear Information System (INIS)

JENSEN, M.A.

2005-01-01

In August 2004, two sealed canisters containing spent nuclear fuel were opened for processing at the Hanford Site's K West fuel storage basin. The fuel was to be processed through cleaning and sorting stations, repackaged into special baskets, placed into a cask, and removed from the basin for further processing and eventual dry storage. The canisters were expected to contain fuel from the old Hanford C Reactor, a graphite-moderated reactor fueled by very low-enriched uranium metal. The expected fuel type was an aluminum-clad slug about eight inches in length and with a weight of about eight pounds. Instead of the expected fuel, the two canisters contained several pieces of thin tubes, some with wire wraps. The material was placed into unsealed canisters for storage and to await further evaluation. Videotapes and still photographs of the items were examined in consultation with available retired Hanford employees. It was determined that the items had a fair probability of being cut-up pieces of fuel rods from the retired Hanford Plutonium Recycle Test Reactor (PRTR). Because the items had been safely handled several times, it was apparent that a criticality safety hazard did not exist when handling the material by itself, but it was necessary to determine if a hazard existed when combining the material with other known types of spent nuclear fuel. Because the PRTR operated more than 40 years ago, investigators had to rely on a combination of researching archived documents, and utilizing common-sense estimates coupled with bounding assumptions, to determine that the fuel items could be handled safely with other spent nuclear fuel in the storage basin. As older DOE facilities across the nation are shut down and cleaned out, the potential for more discoveries of this nature is increasing. As in this case, it is likely that only incomplete records will exist and that it will be increasingly difficult to immediately characterize the nature of the suspect fissionable

On-site concrete cask storage system for spent nuclear fuel

International Nuclear Information System (INIS)

Craig, P.A.; Haelsig, R.T.; Kent, J.D.; Schmoker, D.S.

1989-01-01

A method is described of storing spent nuclear fuel assemblies including the steps of: transferring the fuel assemblies from a spent-fuel pool to a moveable concrete storage cask located outside the spent-fuel pool; maintaining a barrier between the fuel and the concrete in the cask to prevent contamination of the concrete by the fuel; maintaining the concrete storage cask containing the spent-fuel on site at the reactor complex for some predetermined period; transferring the fuel assemblies from the concrete storage cask to a shipping container; and, recycling the concrete storage cask

Fuel element

International Nuclear Information System (INIS)

Hirose, Yasuo.

1982-01-01

Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

The nuclear fuel cycle: (2) fuel element manufacture

International Nuclear Information System (INIS)

Doran, J.

1976-01-01

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

Boiling burnout in the inner annulus of the MK-IV fuel configuration containing ''W'' spring supports

International Nuclear Information System (INIS)

McSweeney, T.I.; Thorne, W.L.; Fitzsimmons, D.E.; Anderson, J.K.

1978-09-01

The establishment of reactor power limits for the NPR is based, to some extent, on the burnout heat transfer results obtained from electrically heated hydraulic models of the reactor coolant passages. Past tests have shown that the outer surface of the middle annulus of the MK-IV fuel configuration goes into burnout before any other surface. Past models have contained no supports in the region where burnout has occurred, yet the reactor configuration must contain supports. The primary purpose of this study is to determine the support influence on the location and magnitude of the burnout heat flux. A second purpose is to establish burnout limits at higher coolant enthalpies. A 2-foot long electrically heated model of the coolant passage, containing supports, has been tested in the high pressure loop of the Thermal Hydraulics Laboratory. Although several supports were located in the active region, a ''W'' spring support was placed in at the channel location where burnout occurred in previous test sections. Thus, the influence of this support has been determined experimentally. Analytically, it has been possible to extend this information to other possible configurations at low quality. A suggested method for using these initial high quality data in the setting of the reactor operating limits is presented

Fuel Handbook[Wood and other renewable fuels

Energy Technology Data Exchange (ETDEWEB)

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

2006-03-15

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

Methanol fuel processor and PEM fuel cell modeling for mobile application

Energy Technology Data Exchange (ETDEWEB)

Chrenko, Daniela [ISAT, University of Burgundy, Rue Mlle Bourgoise, 58000 Nevers (France); Gao, Fei; Blunier, Benjamin; Bouquain, David; Miraoui, Abdellatif [Transport and Systems Laboratory (SeT) - EA 3317/UTBM, Fuel cell Laboratory (FCLAB), University of Technology of Belfort-Montbeliard, Rue Thierry Mieg 90010, Belfort Cedex (France)

2010-07-15

The use of hydrocarbon fed fuel cell systems including a fuel processor can be an entry market for this emerging technology avoiding the problem of hydrogen infrastructure. This article presents a 1 kW low temperature PEM fuel cell system with fuel processor, the system is fueled by a mixture of methanol and water that is converted into hydrogen rich gas using a steam reformer. A complete system model including a fluidic fuel processor model containing evaporation, steam reformer, hydrogen filter, combustion, as well as a multi-domain fuel cell model is introduced. Experiments are performed with an IDATECH FCS1200 trademark fuel cell system. The results of modeling and experimentation show good results, namely with regard to fuel cell current and voltage as well as hydrogen production and pressure. The system is auto sufficient and shows an efficiency of 25.12%. The presented work is a step towards a complete system model, needed to develop a well adapted system control assuring optimized system efficiency. (author)

Fuel exchanging machine for a nuclear ship

International Nuclear Information System (INIS)

Hayashi, Tetsuji.

1984-01-01

Purpose: To prevent atmospheric contaminations upon fuel exchange thereby keep the environmental circumstance clean in the periphery of the nuclear ship. Constitution: A nuclear reactor container is disposed to the inside of a containing vessel in the ship body and a shutter is mounted to the upper opening of the ship body. Further, a landing container having a bottom opening equipped with shutter for alingning the upper opening equipped with shuuter of the ship is elevatably suspended to the trolley of a crane by way of a wire rope and a winch, and a fuel exchange cask is elevatably disposed to the inside of the landing container. Further, airs in the inside of the container is adapted to be discharged externally through a filter by means of a blower and the inside is kept at a negative pressure. Thus, since the containing vessel is covered with the landing container upon fuel exchanging operation, atmospheric contamination can be prevented sufficiently. (Sekiya, K.)

Degradation of EBR-II driver fuel during wet storage

International Nuclear Information System (INIS)

Pahl, R. G.

2000-01-01

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

High density fuels using dispersion and monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

2017-07-01

Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

High density fuels using dispersion and monolithic fuel

International Nuclear Information System (INIS)

Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

2017-01-01

Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

Method to improve lubricity of low-sulfur diesel and gasoline fuels

Science.gov (United States)

Erdemir, Ali

2004-08-31

A method for providing lubricity in fuels and lubricants includes adding a boron compound to a fuel or lubricant to provide a boron-containing fuel or lubricant. The fuel or lubricant may contain a boron compound at a concentration between about 30 ppm and about 3,000 ppm and a sulfur concentration of less than about 500 ppm. A method of powering an engine to minimize wear, by burning a fuel containing boron compounds. The boron compounds include compound that provide boric acid and/or BO.sub.3 ions or monomers to the fuel or lubricant.

Criticality evaluation of long term for spent fuel, using Scale

International Nuclear Information System (INIS)

Esquivel E, J.; Vargas E, S.; Ramirez S, J. R.

2013-10-01

Once carried out the spent fuel discharge, of the reactor core, this continues generating decay heat and diverse fission products, reason why is important to store this fuel inside containers able to dissipate the heat generated by the isotopes decay of the fuel and to maintain the fuels arrangement in subcritical condition. This means that: is necessary to assure the sub-criticality of those fuel assemblies in the time. This work, presents a criticality evaluation of fuel assemblies type PWR in a storage generic container. For this purpose have been used two codes: GeeWiz, to carry out the geometric model of the container with the fuel assemblies, and Keno, with which, the criticality of the full container with fuel is determined until a 10 6 years period. These codes are part of the package Scale. The specifications for each one of the analyzed components are based on a Benchmark document of the Nea/OECD, of where, the results that reports are compared with the obtained results by the realized analysis. (Author)

Explosive composition containing water

Energy Technology Data Exchange (ETDEWEB)

Cattermole, G.R.; Lyerly, W.M.; Cummings, A.M.

1971-11-26

This addition to Fr. 1,583,223, issued 31 May 1968, describes an explosive composition containing a water in oil emulsion. The composition contains an oxidizing mineral salt, a nitrate base salt as sensitizer, water, an organic fuel, a lipophilic emulsifier, and incorporates gas bubbles. The composition has a performance which is improved over and above the original patent.

Improvements in containers

International Nuclear Information System (INIS)

Guy, R.

1977-01-01

An improved container is described for transporting radioactive materials, such as irradiated Magnox fuel elements. It has a lid fixed to the container body and at the corners of the lid has shock absorbers that project from the corners and have part-spheroidal shape. The centre of curvature of the surface of the spheroid is positioned within the lid, so that impact loads on a shock absorber tend to hold it to the container rather than dislodge it. The shock absorbers may be Al-Si alloy castings. (U.K.)

Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface

International Nuclear Information System (INIS)

Enos, David; Bryan, Charles R.

2015-01-01

Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface.

Energy Technology Data Exchange (ETDEWEB)

Enos, David; Bryan, Charles R.

2015-10-01

Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

Alternatives to traditional transportation fuels 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-01

This report provides information on transportation fuels other than gasoline and diesel, and the vehicles that use these fuels. The Energy Information Administration (EIA) provides this information to support the U.S. Department of Energy`s reporting obligations under Section 503 of the Energy Policy Act of 1992 (EPACT). The principal information contained in this report includes historical and year-ahead estimates of the following: (1) the number and type of alterative-fueled vehicles (AFV`s) in use; (2) the consumption of alternative transportation fuels and {open_quotes}replacement fuels{close_quotes}; and (3) the number and type of alterative-fueled vehicles made available in the current and following years. In addition, the report contains some material on special topics. The appendices include a discussion of the methodology used to develop the estimates (Appendix A), a map defining geographic regions used, and a list of AFV suppliers.

Safety issues of dry fuel storage at RSWF

International Nuclear Information System (INIS)

Clarksean, R.L.; Zahn, T.P.

1995-01-01

Safety issues associated with the dry storage of EBR-II spent fuel are presented and discussed. The containers for the fuel have been designed to prevent a leak of fission gases to the environment. The storage system has four barriers for the fission gases. These barriers are the fuel cladding, an inner container, an outer container, and the liner at the RSWF. Analysis has shown that the probability of a leak to the environment is much less than 10 -6 per year, indicating that such an event is not considered credible. A drop accident, excessive thermal loads, criticality, and possible failure modes of the containers are also addressed

Collective radiation doses following a hypothetical, very severe accident to an irradiated fuel transport flask containing AGR fuel

International Nuclear Information System (INIS)

Corbett, J.O.

1985-05-01

Studies of the consequences of very severe, although unlikely, accidents to irradiated fuel transport flasks are made in order to evaluate risks. If an irradiated fuel transport flask carrying AGR fuel were damaged in a hypothetical accident involving a severe impact followed by a prolonged fire, a small proportion of caesium and other fission products might be released to the atmosphere from the gap inventory of broken fuel pins. The consequent radiation dose to the public would arise predominantly by direct irradiation from ground deposits and the ingestion of slightly contaminated foodstuffs. Although these collective doses must generally be estimated with the aid of computer codes, it is shown here that the worst case, when a high proportion of the radioactivity is deposited in a densely population area, can be assessed approximately by a much simpler method, an approach which is of great value in explaining the calculation in a manner that can be readily understood. A comparison is made between the simple approach and equivalent results from the NECTAR code, the worst case is compared with an ensemble average over all weather conditions, and the relative contributions of the two main routes to collective dose are discussed. (author)

MOX fuel fabrication at AECL

International Nuclear Information System (INIS)

Dimayuga, F.C.; Jeffs, A.T.

1995-01-01

Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is nearing completion. This project is funded by the CANDU Owners' Group (COG). The initial fabrication campaign will consist of the production of thirty-eight 37-element (U,Pu)O 2 bundles containing 0.2 wt% Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of the Rehabilitation Project will be given. (author)

Fuel assembly storage pool

International Nuclear Information System (INIS)

Hiranuma, Hiroshi.

1976-01-01

Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

Properties of palm oil fuel ash cement sand brick containing pulverized cockle shell as partial sand replacement

Science.gov (United States)

Mat Aris, S.; Muthusamy, K.; Uzer, A.; Ahmad, S. Wan

2018-04-01

Environmental pollution caused by the disposal of solid wastes generated from both palm oil industry and cockle shell trade has motivated researches to explore the potential of these wastes. Integrating these wastes in production of construction material is one of the ways to reduce amount of waste thrown at dumping area. Thus, the present investigation investigates the performance of palm oil fuel ash (POFA) cement sand brick containing pulverized cockle shell as partial fine aggregate replacement. All mixes used contain 20% of POFA as partial cement replacement. Total of six mixes were prepared by adding a range of pulverized cockle shell that is 0%, 10%, 20%, 30%, 40% and 50% as partial sand replacement. The mixes were prepared in form of brick. All the water cured samples were tested for compressive strength and flexural strength until 28 days. Findings show that brick produced using 20% pulverized cockle shell exhibit the highest compressive strength and flexural strength also the lowest water absorption value.

Fuel Economy Label and CAFE Data

Data.gov (United States)

U.S. Environmental Protection Agency — The Engine and Vehicle Compliance Certification and Fuel Economy Inventory contains measured emissions and fuel economy compliance information for light duty...

Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

International Nuclear Information System (INIS)

Lindemer, T.B.

2002-01-01

Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed

Nuclear fuel assemblies

International Nuclear Information System (INIS)

Butterfield, R.S.; Garner, D.L.M.

1977-01-01

Reference is made to nuclear fuel assemblies designed for cooling on the 'tube-in-shell' principle in which the fuel is contained by a shell and is cooled by coolant passed through tubes extending through the shell. It has been proposed to employ coated particle fuel as a porous bed on the tube side and the bleed coolant from the tubes into direct contact with the fuel particles. In this way heat is extracted both by direct contact with the fuel and by heat transfer through the coolant tube walls. The system described aims to provide an improved structure of tube and shell for a fuel assembly of this kind and is particularly suitable for use in a gas cooled fast reactor, being able to withstand the neutron flux and high temperature conditions in these reactors. Constructional details are given. (U.K.)

Nuclear fuel element

International Nuclear Information System (INIS)

Knowles, A.N.

1979-01-01

A nuclear fuel-containing body for a high temperature gas cooled nuclear reactor is described which comprises a flat plate in which the nuclear fuel is contained as a dispersion of fission product-retaining coated fuel particles in a flat sheet of graphitic or carbonaceous matrix material. The flat sheet is clad with a relatively thin layer of unfuelled graphite bonded to the sheet by being formed initially from a number of separate preformed graphitic artefacts and then platen-pressed on to the exterior surfaces of the flat sheet, both the matrix material and the artefacts being in a green state, to enclose the sheet. A number of such flat plates are supported edge-on to the coolant flow in the bore of a tube made of neutron moderating material. Where a number of tiers of plates are superimposed on one another, the abutting edges are chamfered to reduce vibration. (author)

Canadian fuel development program

International Nuclear Information System (INIS)

Gacesa, M.; Young, E.G.

1992-11-01

CANDU power reactor fuel has demonstrated an enviable operational record. More than 99.9% of the bundles irradiated have provided defect-free service. Defect excursions are responsible for the majority of reported defects. In some cases research and development effort is necessary to resolve these problems. In addition, development initiatives are also directed at improvements of the current design or reduction of fueling cost. The majority of the funding for this effort has been provided by COG (CANDU Owners' Group) over the past 10 to 15 years. This paper contains an overview of some key fuel technology programs within COG. The CANDU reactor is unique among the world's power reactors in its flexibility and its ability to use a number of different fuel cycles. An active program of analysis and development, to demonstrate the viability of different fuel cycles in CANDU, has been funded by AECL in parallel with the work on the natural uranium cycle. Market forces and advances in technology have obliged us to reassess and refocus some parts of our effort in this area, and significant success has been achieved in integrating all the Canadian efforts in this area. This paper contains a brief summary of some key components of the advanced fuel cycle program. (Author) 4 figs., tab., 18 refs

Fueling method in LMFBR type reactors

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Inoue, Kotaro.

1985-01-01

Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

Nuclear fuel pellet loading apparatus

International Nuclear Information System (INIS)

Gerkey, K.S.

1979-01-01

An automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor is described. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

International Nuclear Information System (INIS)

Monteleone, S.

1997-01-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1997-01-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Safeguards approach for irradiated fuel

International Nuclear Information System (INIS)

Harms, N.L.; Roberts, F.P.

1987-03-01

IAEA verification of irradiated fuel has become more complicated because of the introduction of variations in what was once presumed to be a straightforward flow of fuel from reactors to reprocessing plants, with subsequent dissolution. These variations include fuel element disassembly and reassembly, rod consolidation, double-tiering of fuel assemblies in reactor pools, long term wet and dry storage, and use of fuel element containers. This paper reviews future patterns for the transfer and storage of irradiated LWR fuel and discusses appropriate safeguards approaches for at-reactor storage, reprocessing plant headend, independent wet storage, and independent dry storage facilities

A natural-gas fuel processor for a residential fuel cell system

Science.gov (United States)

Adachi, H.; Ahmed, S.; Lee, S. H. D.; Papadias, D.; Ahluwalia, R. K.; Bendert, J. C.; Kanner, S. A.; Yamazaki, Y.

A system model was used to develop an autothermal reforming fuel processor to meet the targets of 80% efficiency (higher heating value) and start-up energy consumption of less than 500 kJ when operated as part of a 1-kWe natural-gas fueled fuel cell system for cogeneration of heat and power. The key catalytic reactors of the fuel processor - namely the autothermal reformer, a two-stage water gas shift reactor and a preferential oxidation reactor - were configured and tested in a breadboard apparatus. Experimental results demonstrated a reformate containing ∼48% hydrogen (on a dry basis and with pure methane as fuel) and less than 5 ppm CO. The effects of steam-to-carbon and part load operations were explored.

Romanian nuclear fuel fabrication and in-reactor fuel operational experience

International Nuclear Information System (INIS)

Budan, O.

2003-01-01

A review of the Romanian nuclear program since mid 60's is made. After 1990, the new Romanian nuclear power authority, RENEL-GEN, elaborated a realistic Nuclear Fuel Program. This program went through the Romanian nuclear fuel plant qualification with the Canadian (AECL and ZPI) support, restarting in January 1995 of the industrial nuclear fuel production, quality evaluation of the fuel produced before 1990 and the recovery of this fuel. This new policy produced good results. FCN is since 1995 the only CANDU fuel supplier from outside Canada recognised by AECL as an authorised CANDU fuel manufacturer. The in-reactor performances and behaviour of the fuel manufactured by FCN after its qualification have been excellent. Very low - more then five times lesser than the design value - fuel defect rate has been recorded up to now and the average discharge of this fuel was with about 9% greater than the design value. Since mid 1998 when SNN took charge of the production of nuclear generated electricity, FCN made significant progresses in development and procurement of new and more efficient equipment and is now very close to double its fuel production capacity. After the completion of the recovery of the fuel produced before June 1990, FCN is already prepared to shift its fuel production to the so-called 'heavy' bundle containing about 19.3 kg of Uranium per bundle

Nuclear fuel string assembly

International Nuclear Information System (INIS)

Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

1976-01-01

A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

Hydrogen fuel cell engines and related technologies

Science.gov (United States)

2001-12-01

The manual documents the first training course developed on the use of hydrogen fuel cells in transportation. The manual contains eleven modules covering hydrogen properties, use and safety; fuel cell technology and its systems, fuel cell engine desi...

Nuclear fuel element, and method of producing same

International Nuclear Information System (INIS)

Armijo, J.S.; Esch, E.L.

1986-01-01

This invention relates to an improvement in nuclear fuel elements having a composite container comprising a cladding sheath provided with a protective barrier of zirconium metal covering the inner surface of the sheath, rendering such fuel elements more resistant to hydrogen accumulation in service. The invention specifically comprises removing substantially all zirconium metal of the barrier layer from the part of the sheath surrounding and defining the plenum region. Thus the protective barrier of zirconium metal covers only the inner surface of the fuel container in the area immediately embracing the fissionable fuel material

Water Reactor Fuel Performance Meeting 2008

International Nuclear Information System (INIS)

2008-10-01

This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

Use of a perfume composition as a fuel for internal combustion engines

NARCIS (Netherlands)

2013-01-01

The present invention relates to fuel compositions containing perfume fractions, that is to say compositions of fragrance materials, and to the use of such perfume fractions containing fuel compositions to provide a fuel for internal combustion engines and burners. According to the present fuel

Increasing TRIGA fuel lifetime with 12 wt.% U TRIGA fuel

Energy Technology Data Exchange (ETDEWEB)

Naughton, W F; Cenko, M J; Levine, S H; Witzig, W F [Pennsylvania State University (United States)

1974-07-01

In-core fuel management studies have been performed for the Penn State Breazeale Reactor (PSBR) wherein 12 wt % U fuel elements are used to replace the standard 8.5 wt % U TRIGA fuel. The core configuration used to develop a calculational model was a 90-element hexagonal array, which is representative of the PSBR core, and consists of five hexagonal rings surrounding a central thimble containing water. The technique employed for refueling the core fully loaded with 8.5 wt % U fuel involves replacing 8.5 wt % U fuel with 12 wt % U fuel using an in-out reloading scheme. A batch reload consists of 6 new 12 wt % U fuel elements. Placing the 12 wt % U fuel in the B ring produces fuel temperatures ({approx}450 {sup o}C) that are well below the 800{sup o}C maximum limitation when the PSBR is operating at its maximum allowed power of 1 Megawatt. The advantages of using new 12 wt % U fuel to replace the burned up 8.5 wt % U fuel in the B ring over refueling strictly with 8.5 wt % U-Zr TRIGA fuel are clearly delineated in Table 1 where cost calculations used the General Atomic pre-1972 prices for TRIGA fuel, i.e., $1500 and $1650 for an 8.5 and 12 wt % U fuel element, respectively. Experimental results obtained to date utilizing the 12 wt % U fuel elements agree with the computed results. (author)

Regulatory viewpoint on nuclear fuel quality assurance

International Nuclear Information System (INIS)

Tripp, L.E.

1976-01-01

Considerations of the importance of fuel quality and performance to nuclear safety, ''as low reasonably achievable'' release of radioactive materials in reactor effluents, and past fuel performance problems demonstrate the need for strong regulatory input, review and inspection of nuclear fuel quality assurance programs at all levels. Such a regulatory program is being applied in the United States of America by the US Nuclear Regulatory Commission. Quality assurance requirements are contained within government regulations. Guidance on acceptable methods of implementing portions of the quality assurance program is contained within Regulatory Guides and other NRC documents. Fuel supplier quality assurance program descriptions are reviewed as a part of the reactor licensing process. Inspections of reactor licensee control of their fuel vendors as well as direct inspections of fuel vendor quality assurance programs are conducted on a regularly scheduled basis. (author)

Technical assistance to AECL: electron beam welding of thick-walled copper containers for nuclear fuel waste disposal

International Nuclear Information System (INIS)

Maak, P.Y.Y.

1984-01-01

This report describes the results of Phase Two of the copper electron beam welding project for the final closure of copper containers for nuclear fuel waste disposal. It has been demonstrated that single pass, electron beam square butt welds (depth of weld penetration > 25 mm) can be made without preheat in both electrolytic tough-pitch copper and oxygen-free copper plates. The present results show that oxygen-free copper exhibits better weldability than the electrolytic tough-pitch copper in terms of weld penetration and vulnerability to weld defects such as gas porosity, erratic metal overflow and blow holes. The results of ultrasonic inspection studies of the welds are also discussed

Development of the scientific concept of the phosphate methods for actinide-containing waste handling (pyrochemical fuel reprocessing)

International Nuclear Information System (INIS)

Orlova, A.I.; Orlova, V.A.; Skiba, O.V.; Bychkov, A.V.; Volkov, Yu.F.; Lukinykh, A.N.; Tomilin, S.V.; Lizin, A.A.

2008-01-01

Full text of publication follows: The crystallochemical phosphate concept in question is developed successfully in the new pyro-electrochemical reprocessing technology of irradiated fuel in molten chlorides of alkaline elements at one of the leading scientific nuclear centers - Research Institute of Atomic Reactors. Irradiated fuel is dissolved in molten chlorides of alkaline elements by mean of treating by chlorine. Then uranium and plutonium dioxides are removed electrochemically. The melt, when used many times, is contaminated by the residual actinide and contains fission products and the so called 'process' elements. This melt is unacceptable for future use. Phosphate methods can be applied for the solution of the following tasks: a) reprocessing (purification) of molten chloride salt solvents; b) conversion of the spent chloride melts to the insoluble stable crystalline product for safe storage and disposal. Within the framework of task 'a' phosphate methods may be realized by the several ways: 1) phosphate concentrating of impurities and their extraction from molten chlorides into solid phase by mean of chemical precipitation, co-precipitation, ion exchange and other chemical interactions, 2) conversion of precipitated waste phosphates to stable crystalline phosphate powders or ceramics for safe storage and disposal. (authors)

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

International Nuclear Information System (INIS)

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

1977-01-01

The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)

Partial oxidation of jet fuels over Rh/Al_2O_3. Design and reaction kinetics of sulfur-containing surrogates

International Nuclear Information System (INIS)

Baer, Julian Nicolaas

2016-01-01

The conversion of logistic fuels via catalytic partial oxidation (CPOX) on Rh/Al_2O_3 at short contact times is an efficient method for generating hydrogen-rich synthesis gas. Depending on the inlet conditions, fuel, and catalyst, high syngas yields, low by-product formation, and rates of high fuel conversion can be achieved. CPOX is relevant for mobile hydrogen generation, e.g., on board of airplanes in order to increase the fuel efficiency via fuel cell-based auxiliary power units. Jet fuels contain hundreds of different hydrocarbons and a significant amount of sulfur. The hydrocarbon composition and sulfur content of a jet fuel vary depending on distributor, origin, and refinement of the crude oil. Little is known about the influence of the various compounds on the synthesis-gas yield and the impact of sulfur on the product yield. In this work, the influence of three main chemical compounds of a jet fuel (aromatics, alkanes, and sulfur compounds) on syngas selectivity, the catalyst deactivation process, and reaction sequence is unraveled. As representative components of alkanes and aromatics, n-dodecane and 1,2,4-trimethylbenzene were chosen for ex-situ and in-situ investigations on the CPOX over Rh/Al_2O_3, respectively. Additionally, for a fixed paraffin-to-aromatics ratio, benzothiophene or dibenzothiophene were added as a sulfur component in three different concentrations. The knowledge gained about the catalytic partial oxidation of jet fuels and their surrogates is used to identify requirements for jet fuels in mobile applications based on CPOX and to optimize the overall system efficiency. The results show an influence of the surrogate composition on syngas selectivity. The tendency for syngas formation increases with higher paraffin contents. A growing tendency for by-product formation can be observed with increasing aromatics contents in the fuel. The impact of sulfur on the reaction system shows an immediate change in the product distribution. An

System and method for consolidating spent fuel rods

International Nuclear Information System (INIS)

Baudro, T.O.

1987-01-01

A system is described for consolidating spent fuel rods from spent fuel assemblies, comprising: a consolidation container in which the fuel rods may be packed; a frame capable of holding a fuel assembly and the container during consolidation, the frame permitting each of the fuel assembly and the container to be removed; tool means with gripper means for gripping and releasing a rod, the tool means including means for moving the gripper means upwardly and downwardly; a first indexing head having first guide means for guiding the gripper means while the gripper means moves downwardly; a first rail, the first indexing head being slidably mounted on the first rail; a second indexing head having second guide means for guiding the gripper means while the gripper means moves downwardly; a second rail, the second indexing head being slidably mounted on the second rail; and a third rail, the first rail and the second rail being slidably mounted on the third rail; wherein the first indexing head is slidable on the first and third rails to a first position that is above a preselected rod in the fuel assembly; and wherein the second indexing head is slidable on the second and third rails to a second position that is above a preselected location in the container

Characterization of graphite-matrix pulsed reactor fuels

International Nuclear Information System (INIS)

Karnes, C.H.; Marion, R.H.

1976-01-01

The performance of the Annular Core Pulsed Reactor (ACPR) is being upgraded in order to accommodate higher fluence experiments for fast reactor fuel element transient and safety studies. The increased fluence requires a two-zone core with the inner zone containing fuel having a high enthalpy and the capability of withstanding very high temperatures during both pulsed and steady state operation. Because the fuel is subjected to a temperature risetime of 2 to 5 ms and to a large temperature difference across the diameter, fracture due to thermal stresses is the primary failure mode. One of the fuels considered for the high enthalpy inner region is a graphite-matrix fuel containing a dispersion of uranium--zirconium carbide solid solution particles. A program was initiated to optimize the development of this class of fuel. This summary presents results on formulations of fuel which have been fabricated by the Materials Technology Group of the Los Alamos Scientific Laboratory

Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

International Nuclear Information System (INIS)

Sonoda, Yukio; Tamaoki, Tetsuo

1989-01-01

Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

Fuel assembly

International Nuclear Information System (INIS)

Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

1998-01-01

In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

Status of LEU fuel development and conversion of NRU

International Nuclear Information System (INIS)

Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

1989-11-01

The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U 3 Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U 3 Si 2 dispersion fuel to complement out Al-U 3 Si capability. Three full-size NRU rods containing Al-U 3 Si 2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U 3 Si 2 fuel revealed that the U 3 Si 2 behaved similarly to U 3 Si 2 fuel revealed that the U 3 Si 2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

Estimate of the crud contribution to shipping cask containment requirements

International Nuclear Information System (INIS)

Sandoval, R.P.; Einziger, R.E.; Jordan, H.; Malinauskas, A.P.; Mings, W.J.

1992-01-01

This paper reports that a methodology is developed to relate U.S. Code of Federal Regulations, Title 10, Part 71 (10CFR71) containment requirements to leak rates for the special case in which the only radioactive species having a potential for escape form the cask is that associated with debris (crud) contained on the fuel assemblies being transported. The methodology accounts for the characteristics of the crud and for attenuation of the gas-borne crud particulates once they become suspended within the cask. Calculations are performed for typical spent-fuel transport cask geometries and the normal and accident conditions prescribed in 10CFR71. The most current published data are used for crud composition and structure, specific activity, spallation mechanics and fractions, and crud particle size. The containment criteria leak rates are calculated assuming 5-yr-old spent fuel. In each accident case, the containment leak rate criteria are well in excess of 10 cm 3 /s. Under normal conditions of transport, the regulatory containment requirements are met by leak rates ranging from 1.5 x 10 -3 cm 3 /s to 1.5 x 10 -4 cm 3 /s for the transport of boiling water reactor fuel assemblies and form 1.8 x 10 -2 cm 3 /s to 1.3 x 10 -3 cm 3 /s for pressurized water reactor fuel assemblies. The calculated leak rates are most sensitive to the cask design, type of fuel, and particle size distribution. Conservatism of the limiting leak rates is discussed

Development of the CANDU high-burnup fuel design/analysis technology

Energy Technology Data Exchange (ETDEWEB)

Suk, Ho Chun; Sim, K. S.; Oh, D. J.; Park, J. H.; Jun, J. S.; Yoo, K. J.

1997-08-01

This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs.

Development of the CANDU high-burnup fuel design/analysis technology

International Nuclear Information System (INIS)

Suk, Ho Chun; Sim, K. S.; Oh, D. J.; Park, J. H.; Jun, J. S.; Yoo, K. J.

1997-08-01

This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs

Detection method of a failed fuel

International Nuclear Information System (INIS)

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

1976-01-01

Object: To divide a tank arrangement into a heating tank for the exclusive use of heating and a mixing tank for the exclusive use of mixing to thereby minimize the purifying amount of reactor water pumped from the interior of reactor and to considerably minimize the capacity of a purifier. Structure: In a detection method of a failed fuel comprising stopping a flow of coolant within fuel assemblies arranged in the coolant in a reactor container, sampling said coolant within the fuel assemblies, and detecting a radioactivity level of sampling liquid, the improvement of the method comprising the steps of heating a part of said coolant removed from the interior of said reactor container, mixing said heated coolant into the remainder of said removed coolant, pouring said mixed liquid into said fuel assemblies, and after a lapse of given time, sampling the liquid poured into said fuel assemblies. (Kawakami, Y.)

Temperature behavior of 12 wt.% U TRIGA fuel

Energy Technology Data Exchange (ETDEWEB)

Levine, S H; Geisler, G C; Totenbier, R E [Pennsylvania State University (United States)

1974-07-01

Stainless steel clad 12 wt % U TRIGA fuel elements have been used to refuel the Penn State University's Breazeale Reactor (PSBR). When 12 wt % U fuel containing nominally 55 gms of {sup 235}U per fuel element is substituted for the 8.5 wt % U fuel containing nominally 38 gms {sup 235}U, higher fuel temperatures were produced in the 12 wt % U fuel than in the 8.5 wt % U fuel at the same reactor powers. The higher fuel temperature can be related to the higher power densities in the 12 wt % U fuel. The power density is calculated to be 35% higher in the 12 wt % U fuel when 6 of these fuel elements are substituted for 8.5 wt % U fuel in the innermost ring, the B ring. Temperatures have been calculated for the 12 wt % U fuel in the above configuration for both steady state and pulse conditions, assuming a 35% higher fuel density in the 12 wt % U fuel and the results compare favorably with the experimental measurements. This is particularly true when the comparison is made with temperature data taken after exposing the new fuel elements to a series of pulses. These calculations and data will be presented at the meeting. (author)

Status of work on the final repository concept concerning direct disposal of spent fuel rods in fuel rod casks (BSK)

International Nuclear Information System (INIS)

Filbert, W.; Wehrmann, J.; Bollingerfehr, W.; Graf, R.; Fopp, S.

2008-01-01

The reference concept in Germany on direct final storage of spent fuel rods is the burial of POLLUX containers in the final repository salt dome. The POLLUX container is self-shielded. The final storage concept also includes un-shielded borehole storage of high-level waste and packages of compacted waste. GNS has developed a spent fuel container (BSK-3) for unshielded borehole storage with a mass of 5.2 tons that can carry the fuel rods of three PWR reactors of 9 BWR reactors. The advantages of BSK storage include space saving, faster storage processes, less requirements concerning technical barriers, cost savings for self-shielded casks.

Gas flow in and out of a nuclear waste container

International Nuclear Information System (INIS)

Zwahlen, E.D.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

1989-05-01

We analyze the flow of gases out of and into a high-level-waste container in the unsaturated tuff of Yucca Mountain. Containers are expected to fail eventually by localized cracks and penetrations. Even though the penetrations may be small, argon gas initially in the hot container can leak out. As the waste package cools, the pressure inside the container can become less than atmospheric, and air can leak in. 14 C released from the hot fuel-cladding surface can leak out of penetrations, and air inleakage can mobilize additional 14 C and other volatile radioactive species as it oxidizes the fuel cladding and the spent fuel. In an earlier paper we studied the gas flow through container penetrations occurring at the time of emplacement. Here we analyze the flow of gas for various penetration sizes occurring at 300 years. 3 refs., 2 figs

Rack for nuclear fuel elements

International Nuclear Information System (INIS)

Rubinstein, H.J.; Gordon, C.B.; Robison, A.; Clark, P.M.

1977-01-01

Disclosed is a rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed spent fuel elements. Each fuel element is supported at the lower end thereof by a respective support that rests on the floor of the spent fuel pool for a nuclear power plant. An open rack frame is employed as an upright support for the enclosures containing the spent fuel elements. Legs at the lower corners of the frame rest on the floor of the pool to support the frame. In one exemplary embodiment, the support for the fuel element is in the form of a base on which a fuel element rests and the base is supported by legs. In another exemplary embodiment, each fuel element is supported on the pool floor by a self-adjusting support in the form of a base on which a fuel element rests and the base rests on a ball or swivel joint for self-alignment. The lower four corners of the frame are supported by legs adjustable in height for leveling the frame. Each adjustable frame leg is in the form of a base resting on the pool floor and the base supports a threaded post. The threaded post adjustably engages a threaded column on which rests the lower end of the frame. 16 claims, 14 figures

Fuel Economy Label and CAFE Data

Science.gov (United States)

The Engine and Vehicle Compliance Certification and Fuel Economy Inventory contains measured emissions and fuel economy compliance information for light duty vehicles. Data is collected by EPA to certify compliance with the applicable fuel economy provisions of the Energy Policy and Conservation Act (EPCA) and The Energy Independent Security Act of 2007

Nuclear reactor fuel element splitter

International Nuclear Information System (INIS)

Yeo, D.

1976-01-01

A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

Effect of absorbing impurities on the accuracy of the optical method for the detection of the iodine-containing substances resulting from the processing of waste nuclear fuel

Science.gov (United States)

Kireev, S. V.; Simanovsky, I. G.; Shnyrev, S. L.

2010-12-01

The study is aimed at an increase in the accuracy of the optical method for the detection of the iodine-containing substances in technological liquids resulting form the processing of the waste nuclear fuel. It is demonstrated that the accuracy can be increased owing to the measurements at various combinations of wavelengths depending on the concentrations of impurities that are contained in the sample under study and absorb in the spectral range used for the detection of the iodine-containing substances.

Method to mount defect fuel elements i transport casks

International Nuclear Information System (INIS)

Borgers, H.; Deleryd, R.

1996-01-01

Leaching or otherwise failed fuel elements are mounted in special containers that fit into specially designed chambers in a transportation cask for transport to reprocessing or long-time storage. The fuel elements are entered into the container under water in a pool. The interior of the container is dried before transfer to the cask. Before closing the cask, its interior, and the exterior of the container are dried. 2 figs

Spent fuel management newsletter. No. 2

International Nuclear Information System (INIS)

1993-04-01

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

Spent fuel management newsletter. No. 2

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-04-01

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

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E. D.

1984-01-01

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

Nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hindle, E. D.

1984-10-16

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

Nuclear reactor using fuel sphere for combustion and fuel spheres for breeding

International Nuclear Information System (INIS)

Yamashita, Kiyonobu.

1995-01-01

The present invention concerns a pebble bed-type reactor which can efficiently convert parent nuclides to fission nuclides. Fuel spheres for combustion having fission nuclides as main fuels, and fuel spheres for breeding having parent nuclides as main fuels are used separately, in the pebble bed-type reactor. According to the present invention, fuel spheres for breeding can be stayed in a reactor core for a long period of time, so that parent nuclides can be sufficiently converted into fission nuclides. In addition, since fuel spheres for breeding are loaded repeatedly, the amount thereof to be used is reduced. Therefore, the amount of the fuel spheres for breeding is small even when they are re-processed. On the other hand, since the content of the fission nuclides in the fuel spheres for breeding is not great, they can be put to final storage. This is attributable that although the fuel spheres for breeding contain fission nuclides generated by conversion, the fission nuclides are annihilated by nuclear fission reactions at the same time with the generation thereof. (I.S.)

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

ITER fuel cycle systems layout

International Nuclear Information System (INIS)

Kveton, O.K.

1990-10-01

The ITER fuel cycle building (FCB) will contain the following systems: fuel purification - permeator based; fuel purification - molecular sieves; impurity treatment; waste water storage and treatment; isotope separation; waste water tritium extraction; tritium extraction from solid breeder; tritium extraction from test modules; tritium storage, shipping and receiving; tritium laboratory; atmosphere detritiation systems; fuel cycle control centre; tritiated equipment maintenance space; control maintenance space; health physics laboratory; access, access control and facilities. The layout of the FCB and the requirements for these systems are described. (10 figs.)

No significant fuel failures (NSFF)

International Nuclear Information System (INIS)

Domaratzki, Z.

1979-01-01

It has long been recognized that no emergency core cooling system (ECCS) could be absolutely guaranteed to prevent fuel failures. In 1976 the Atomic Energy Control Board decided that the objective for an ECCS should be to prevent fuel failures, but if the objective could not be met it should be shown that the consequences are acceptable for dual failures comprising any LOCA combined with an assumed impairment of containment. Out of the review of the Bruce A plant came the definition of 'no significant fuel failures': for any postulated LOCA combined with any one mode of containment impairment the resultant dose to a person at the edge of the exclusion zone is less than the reference dose limits for dual failures

CONTAINMENT EVALUATION OF BREACHED AL-SNF FOR CASK TRANSPORT

International Nuclear Information System (INIS)

Vinson, D. W.; Sindelar, R. L.; Iyer, N. C.

2005-01-01

Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are ''failed'' or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, L R , for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems. The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A 2 , are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate

Thermal breeder fuel enrichment zoning

International Nuclear Information System (INIS)

Capossela, H.J.; Dwyer, J.R.; Luce, R.G.; McCoy, D.F.; Merriman, F.C.

1992-01-01

A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect. 1 figure

TMI-2 spent fuel shipping

International Nuclear Information System (INIS)

Quinn, G.J.; Burton, H.M.

1985-01-01

TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

Determining fissile content of nuclear fuel elements

International Nuclear Information System (INIS)

Arya, S.P.; Grossman, L.N.; Schoenig, F.C.

1980-01-01

This invention relates to the determination of the fissile fuel content of fuel for nuclear reactors. A nondestructive method is described for determining rapidly, accurately and simultaneously the fissile content, enrichment and location of fuel material which may also contain amounts of burnable poison, by detecting the γ-rays emitted from the fuel material due to natural radioactive decay. (U.K.)

Thermodynamic properties of the DUPIC fuel and its performance

Energy Technology Data Exchange (ETDEWEB)

Park, Kwang Heon; Kim, Hee Moon [Kyung Hee Univ., Seoul (Korea, Republic of)

1997-07-01

This study describes thermodynamic properties of DUPIC fuel and performance. In initial state, DUPIC fuel which contains fissile materials is different from general nuclear fuel. So this study analyzed oxygen potential, thermal conductivity and specific heat of the DUPIC fuel.

Leaching studies of heavy concrete material for nuclear fuel waste immobilization containers

International Nuclear Information System (INIS)

Onofrei, M.; Raine, D.; Brown, L.; Hooton, R.D.

1989-08-01

The leaching behaviour of a high-density concrete was studied as part of a program to evaluate its potential use as a container material for nuclear fuel waste under conditions of deep geologic disposal. Samples of concrete material were leached in deionized distilled water, Standard Canadian Shield Saline Solution (SCSSS), SCSSS plus 20% Na-bentonite, and SCSSS plus granite and 20% Na-bentonite under static conditions at 100 degrees celsius for periods up to 365 days. The results of these leaching experiments suggest that the stability of concrete depends on the possible internal structural changes due to hydration reactions of unhydrated components, leading to the formation of C-S-H gel plus portlandite (Ca(OH) 2 ). The factors controlling the concrete leaching process were the composition of the leachant and the concentration of elements in solution capable of forming precipitates on the concrete surface, e.g., silicon, Mg 2+ and Ca 2+ . The main effect observed during leaching was an increase in groundwater pH (from 7 to 9). However, the addition of Na-bentonite suppressed the normal tendency of the pH of the groundwater in contact with concrete to rise rapidly. It was shown that the solution concentration of elements released from the concrete, particularly potassium, increased in the presence of Na-bentonite

Proceedings of the fuel cells `95 review meeting

Energy Technology Data Exchange (ETDEWEB)

George, T.J.

1995-08-01

This document contains papers presented at the Fuel Cells `95` Review Meeting. Topics included solid oxide fuel cells; DOE`s transportation program; ARPA advanced fuel cell development; molten carbonate fuel cells; and papers presented at a poster session. Individual papers have been processed separately for the U.S. DOE databases.

High polymer composites for containers for the long-term storage of spent nuclear fuel and high level radioactive waste

International Nuclear Information System (INIS)

Bonin, H.W.; Vui, V.T.; Legault, J.-F.

1997-01-01

The feasibility of using polymeric composite materials as an alternative to metals in the design of a nuclear waste disposal container was examined. The disposal containers would be stored in deep underground vaults in plutonic rock formations within the Canadian Shield for several thousands of years. The conditions of disposal considered in the evaluation of the polymeric composite materials were based on the long-term disposal concept proposed by Atomic Energy of Canada Limited. Four different composites were considered for this work, all based on boron fibre as reinforcing material, imbedded in polymeric matrices made of polystyrene (PS), polymethyl methacrylate (PMMA), Devcon 10210 epoxy, and polyetheretherketone (PEEK). Both PS and PMMA were determined as unsuitable for use in the fabrication of the storage container because of thermal failure. This was determined following thermal analysis of the materials in which heat transfer calculations yielded the temperature of the container wall and of the surroundings resulting from the heat generated by the spent nuclear fuel stored inside the container. In the case of the PS, the temperature of the container, the buffer and the backfill would exceed the 100 degrees C imposed in the AECL's proposal as the maximum allowable. In the case of the PMMA, the 100 degrees temperature is too close to the glass transition temperature of this material (105 degrees C) and would cause structural degradation of the container wall. The other two materials present acceptable thermal characteristics for this application. An important concern for polymeric materials in such use is their resistance to radiations. The Devcon 10210 epoxy has been the object of research at the Royal Military College in the past years and fair, but limited, resistance to both neutrons and gamma radiation has been demonstrated, with the evidence of increased mechanical strength when subjected to moderate doses. Provided that the container wall could be

Tritium distribution between the fuel can and the oxide of fuel elements of light-water reactors

International Nuclear Information System (INIS)

Masson, M.

1986-12-01

The study on the measurement of tritium and other radionuclide contained in zircaloy fuel cans of the water cooled reactor fuel elements had two aims: the first was to estimate with accuracy the distribution of tritium in a fuel element (can + oxide). The measurement of tritium in the zircaloy fuel cans of the BORSSELE fuel elements associated with the measurement of tritium in the oxide allowed the establishment of a complete tritium balance on an industrial spent fuel element. This result has been compared to the values calculated by the code CEA/SEN and will allow to validate or adjust this calculation. The second aim delt with the characterization of the other radionuclides gaseous (Kr85) or not (Cs 134 and 137) contained in the solid zircaloy wastes (hulls) coming from the industrial reprocessing of ''water cooled'' fuel elements. These activity measurements in the hulls allowed to estimate the residual content of tritium, Kr 85 and other radionuclides which may be found in these solid wastes (high-level βγ radioactive wastes). Original experimental methods have been developed to reach these aims (dissolution in ammonium bifluoride medium and quantitative recovery of gases produced, radiochromatography, and liquid scintillation after double distillation). One tries to explain the presence of Kr 85 in the irradiated can [fr

Low enrichment fuel conversion for Iowa State University

International Nuclear Information System (INIS)

Rohach, A.F.; Hendrickson, R.A.

1990-08-01

Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. The LEU fuel has not been received. The HEU fuel assemblies for the UTR-10 reactor will not fit into any current research reactor shipping containers; therefore, the fuel assemblies must be disassembled and the fuel shipped as fuel plates. Procedures and practices have been developed so that the fuel assemblies will be disassembled in a shielded environment

Fuel rods

International Nuclear Information System (INIS)

Adachi, Hajime; Ueda, Makoto

1985-01-01

Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

Fission product vapour - aerosol interactions in the containment: simulant fuel studies

International Nuclear Information System (INIS)

Beard, A.M.; Benson, C.G.; Bowsher, B.R.

1988-12-01

Experiments have been conducted in the Falcon facility to study the interaction of fission product vapours released from simulant fuel samples with control rod aerosols. The aerosols generated from both the control rod and fuel sample were chemically distinct and had different deposition characteristics. Extensive interaction was observed between the fission product vapours and the control rod aerosol. The two dominant mechanisms were condensation of the vapours onto the aerosol, and chemical reactions between the two components; sorption phenomena were believed to be only of secondary importance. The interaction of fission product vapours and reactor materials aerosols could have a major impact on the transport characteristics of the radioactive emission from a degrading core. (author)

The Caramel fuel in OSIRIS

International Nuclear Information System (INIS)

Cherruau, Francois.

1980-11-01

This paper presents the main characteristics of the caramel fuel, a description of OSIRIS transformations that were decided in line with its conversion and the results of its operation since then. The Caramel fuel is made from sintered UO 2 pellets contained in zircaloy clads forming the plates of the fuel assembly reducing the enrichment need to as little as 3 to 10% instead of 93% enriched U/Al in the previous fuel. The first year of experience shows the capacity under a statistic scale of the caramel fuel to fulfil the most severe operation requirements for use in low and medium power research reactors

Characterization of the Microstructure of Irradiated U-Mo Dispersion Fuel with a Matrix that Contains Si

International Nuclear Information System (INIS)

Keiser, Jr. D.D.; Robinson, A.B.; Jue, J.F.; Medvedev, P.; Finlay, M.R.

2009-01-01

RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with Al-2Si matrix after irradiation to around 50% LEU burnup. Si-rich layers were observed in many areas around the various U-7Mo fuel particles. In one local area of one of the samples, where the Si-rich layer had developed into a layer devoid of Si, relatively large fission gas bubbles were observed in the interaction phase. There may be a connection between the growth of these bubbles and the amount of Si present in the interaction layer. Overall, it was found that having Si-rich layers around the fuel particles after fuel plate fabrication positively impacted the overall performance of the fuel plate

Evaluation of the criticality of a concrete container for storage of spent fuel in dry with MCNP; Evaluacion de la criticidad de un contenedor de concreto para almacenamiento de combustible gastado en seco con MCNP

Energy Technology Data Exchange (ETDEWEB)

Xolocostli M, J. V.; Ramirez S, J. R., E-mail: vicente.xolocostli@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

A main concern exists inside the nuclear power plants in operation around the world that is the with respect to the storage capacity of the spent fuel, due to the useful life of the plant and the storage capacity in the spent fuel pool. In diverse countries is believed that one of the best alternatives for the spent fuel is the reprocessing of the same one since exists a great quantity of fissile material that can be profitable as the Pu-239, but even so the costs for the reprocessing continue being high, what limits taking this process to great scale. Is for that reason the importance of the containers for storage of spent fuel in dry which has had a great apogee in the last years, since they represent an alternative to store the spent fuel before making a decision on the reprocessing of the same one or the final disposal. In this work an evaluation of the criticality of a concrete container for storage of spent fuel in dry commercially available is made, and which is useful for fuel assemblies type PWR like BWR, in our case only the type BWR is considered. For the analysis of the evaluation was used the code MCNP5, considering the characteristics of the concrete container according to the available data, although the type of fuel assembly is BWR one of the models of the ABB company was considered with which the comparative of the results is made. The made calculations were carried out considering the inundation of the gap that exist and the external cavity, being this the most extreme condition to arrive to the criticality or in the case of happening an accident to have the filtration of the water toward the space of the gap. (author)

An optimized BWR fuel lattice for improved fuel utilization

International Nuclear Information System (INIS)

Bernander, O.; Helmersson, S.; Schoen, C.G.

1984-01-01

Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

Science.gov (United States)

Moore, R.H.

1962-10-01

A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

Automotive fuels - environmental and health implications

International Nuclear Information System (INIS)

Lucas, A.G.

1992-01-01

This document covers papers presented to the Institute of Petroleum's conference ''Automotive Fuels: Environmental and Health Implications'' held on the 9th October 1991. This wide ranging title meant that topics covered included the biochemistry, pathology and epidemiology of automotive fuel use, combustion science, environmental chemistry and atmospheric modelling. Also discussed are the technology of fuel and engine manufacture, limiting and containing emissions and social and political aspects relating to the use of automotive fuels. (UK)

Encapsulation of spent nuclear fuel in ceramic materials

International Nuclear Information System (INIS)

Forberg, S.; Westermark, T.

1983-03-01

The international situation with regard to deposition of spent nuclear fuel is surveyed, with emphasis on encapsulation in ceramic materials. The feasibility and advantages of ceramic containers, thermodynamic stable in groundwater, are discussed as well as the possibility to ensure that stability for longevity by engineered measures. The design prerequisite are summarized and suggestions are made for a conceptual design, comprising rutile containers with stacks of coiled fuel pins. A novel technique is suggested for the homogeneous sealing of rutile containers at low temperatures. acceptable also for the fuel pin package. Key points are given for research, demonstration and verifications of the design foundations and for future improvements. Of which a few ideas are exemplified. (author)

Fuel production for LWRs - MOX fuel aspects

International Nuclear Information System (INIS)

Deramaix, P.

2005-01-01

Plutonium recycling in Light Water Reactors is today an industrial reality. It is recycled in the form of (U, Pu)O 2 fuel pellets (MOX), fabricated to a large extent according to UO 2 technology and pellet design. The similarity of physical, chemical, and neutron properties of both fuels also allows MOX fuel to be burnt in nuclear plants originally designed to burn UO 2 . The industrial processes presently in use or planned are all based on a mechanical blending of UO 2 and PuO 2 powders. To obtain finely dispersed plutonium and to prevent high local concentration of plutonium, the feed materials are micronised. In the BNFL process, the whole (UO 2 , PuO 2 ) blend is micronised by attrition milling. According to the MIMAS process, developed by BELGONUCLEAIRE, a primary blend made of UO 2 containing about 30% PuO 2 is micronised in a ball mill, afterwards this primary blend is mechanically diluted in UO 2 to obtain the specified Pu content. After mixing, the (U, Pu)O 2 powder is pressed and the pellets are sintered. The sintering cover gas contains moisture and 5 v/o H 2 . Moisture increases the sintering process and the U-Pu interdiffusion. After sintering and grinding, the pellets are submitted to severe controls to verify conformity with customer specifications (fissile content, Pu distribution, surface condition, chemical purity, density, microstructure). (author)

Fuel cycle problems in fusion reactors

International Nuclear Information System (INIS)

Hickman, R.G.

1976-01-01

Fuel cycle problems of fusion reactors evolve around the breeding, recovery, containment, and recycling of tritium. These processes are described, and their implications and alternatives are discussed. Technically, fuel cycle problems are solvable; economically, their feasibility is not yet known

STACE: source term analyses for containment evaluations of transport casks

International Nuclear Information System (INIS)

Seager, K.D.; Gianoulakis, S.E.; Barrett, P.R.; Rashid, Y.R.; Reardon, P.C.

1993-01-01

STACE evaluates the calculated fuel rod response against failure criteria based on the cladding residual ductility and fracture properties as functions of irradiation and thermal environments. The fuel rod gap inventory contains three forms of releasable RAM: (1) gaseous, e.g., 85 Kr, (2) volatiles, e.g., 134 Cs and 137 Cs, and (3) actinides associated with fuel fines. The quantities of these products are limited to that contained within the fuel-cladding gap region and associated interconnected voids. Cladding pinhole failure will also result in the ejection of about 0.003 percent of the fuel, in the form of fines, into the cask cavity. Significant attenuation of the aerosol concentration in the transport cask can occur, depending upon the residence time of the aerosol in the cask compared with its rate of escape from the cask into the environment. (J.P.N.)

Gamma radiolysis effects on leaching behavior of ceramic materials for nuclear fuel waste immobilization containers

International Nuclear Information System (INIS)

Onofrei, M.; Raine, D.K.; Hocking, W.H.; George, K.; Betteridge, J.S.

1986-01-01

The leaching behavior of ceramic materials for nuclear fuel waste immobilization containers, under the influence of a moderate gamma dose rate (4 Gy/h), has been investigated. Samples of Al/sub 2/O/sub 3/, stabilized ZrO/sub 2/, TiO/sub 2/, cermet (70% Al/sub 2/O-30% TiC), porcelain (with high Al/sub 2/O/sub 3/ content), and concrete (with sulfate-resisting portland cement plus silica fume) have been leached in Standard Canadian Shield Saline Solution (SCSSS), and SCSSS plus clay and sand (components of the disposal system), at 100 0 and 150 0 C for 231 and 987 days, respectively. Leaching solutions were analyzed and the surfaces of the leached samples were investigated by scanning electron microscopy in conjunction with energy dispersive X-ray spectroscopy and secondary ion mass spectrometry. Radiolysis did not appear to enhance the leaching, with or without bentonite and sand in the system. Analysis of the gas phase from sealed capsules showed O/sub 2/ depletion and production of CO/sub 2/ in all experiments containing bentonite. The decrease in O/sub 2/ is attributed to the leaching from the clay of Fe(II) species, which can participate in redox reactions with radicals generated by radiolysis. The CO/sub 2/ is produced from either the organic or inorganic fraction in the bentonite

Nuclear fuels

International Nuclear Information System (INIS)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

2009-01-01

Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

Nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

2009-07-01

Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

Development of a method for detecting nuclear fuel debris and water leaks at a nuclear reactor/containment vessel by flow visualization

International Nuclear Information System (INIS)

Umezawa, Shuichi; Tanaka, Katsuhiko

2013-01-01

It is the important issue to fill up each nuclear reactor/containment vessel with water and to take out debris of damaged fuel from them for decommissioning of Fukushima Daiichi nuclear power plants. It is necessary to detect the debris and water leaks at a nuclear reactor/containment vessel for the purpose. However, the method is not completely developed in the present stage. Accordingly, we have developed a method for detecting debris and water leaks at a nuclear reactor/containment vessel by flow visualization. Experiments of the flow visualization were conducted using two types of water tanks. An optical fiber and a collimator lens were employed for modifying a straight laser beam into a sheet projection. Some visualized images were obtained through the experiments. Particle Image Velocimetry, i.e. PIV, analysis was applied to the images for quantitative flow rate analysis. Consequently, it is considered that the flow visualization method has a possibility for the practical use. (author)

Fuel quality issues in stationary fuel cell systems.

Energy Technology Data Exchange (ETDEWEB)

Papadias, D.; Ahmed, S.; Kumar, R. (Chemical Sciences and Engineering Division)

2012-02-07

Fuel cell systems are being deployed in stationary applications for the generation of electricity, heat, and hydrogen. These systems use a variety of fuel cell types, ranging from the low temperature polymer electrolyte fuel cell (PEFC) to the high temperature solid oxide fuel cell (SOFC). Depending on the application and location, these systems are being designed to operate on reformate or syngas produced from various fuels that include natural gas, biogas, coal gas, etc. All of these fuels contain species that can potentially damage the fuel cell anode or other unit operations and processes that precede the fuel cell stack. These detrimental effects include loss in performance or durability, and attenuating these effects requires additional components to reduce the impurity concentrations to tolerable levels, if not eliminate the impurity entirely. These impurity management components increase the complexity of the fuel cell system, and they add to the system's capital and operating costs (such as regeneration, replacement and disposal of spent material and maintenance). This project reviewed the public domain information available on the impurities encountered in stationary fuel cell systems, and the effects of the impurities on the fuel cells. A database has been set up that classifies the impurities, especially in renewable fuels, such as landfill gas and anaerobic digester gas. It documents the known deleterious effects on fuel cells, and the maximum allowable concentrations of select impurities suggested by manufacturers and researchers. The literature review helped to identify the impurity removal strategies that are available, and their effectiveness, capacity, and cost. A generic model of a stationary fuel-cell based power plant operating on digester and landfill gas has been developed; it includes a gas processing unit, followed by a fuel cell system. The model includes the key impurity removal steps to enable predictions of impurity breakthrough

Design of fuel loading for Bohunice V-1 Unit 2 reaktor for fuel cycle No.19

International Nuclear Information System (INIS)

Majercik, J.

1998-01-01

The report contains description of the design of fuel loading for the fuel cycle No. 19 in the V-1 Bohunice Unit 2 reactor. Input data and computer codes used for the development of the design are shown. The fuel loading is characterized by the assortment of the fuel loaded and by the scheme of re shuffling of assemblies in the core. An evaluation of basic neutronic core parameters as relates to the compliance with safety criteria is a part of the report as well

Fuel rods

International Nuclear Information System (INIS)

Fukushima, Kimichika.

1984-01-01

Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

Gas-to-liquids synthetic fuels for use in fuel cells : reformability, energy density, and infrastructure compatibility.

Energy Technology Data Exchange (ETDEWEB)

Ahmed, S.; Kopasz, J. P.; Russell, B. J.; Tomlinson, H. L.

1999-09-08

The fuel cell has many potential applications, from power sources for electric hybrid vehicles to small power plants for commercial buildings. The choice of fuel will be critical to the pace of its commercialization. This paper reviews the various liquid fuels being considered as an alternative to direct hydrogen gas for the fuel cell application, presents calculations of the hydrogen and carbon dioxide yields from autothermal reforming of candidate liquid fuels, and reports the product gas composition measured from the autothermal reforming of a synthetic fuel in a micro-reactor. The hydrogen yield for a synthetic paraffin fuel produced by a cobalt-based Fischer-Tropsch process was found to be similar to that of retail gasoline. The advantages of the synthetic fuel are that it contains no contaminants that would poison the fuel cell catalyst, is relatively benign to the environment, and could be transported in the existing fuel distribution system.

Inerting Aircraft Fuel Systems Using Exhaust Gases

Science.gov (United States)