Development of boronated aluminum alloy for basket of cask for nuclear spent fuel

International Nuclear Information System (INIS)

Sakaguchi, Y.; Saida, T.; Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

2001-01-01

Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities and competent authorities in Japan, and have established transport and storage cask design ''MSF series'' which realizes higher payload and reliability for long term storage. MSF series transport and storage cask uses new-developed boronated aluminum as basket material. This boronated aluminum has been developed to improve characteristics of material. To achieve this object, powder metallurgy method has been adopted for manufacturing boronated material. It is well known that this method provides excellent characteristics for the material and this boronated aluminum alloy has obtained excellent both mechanical and neutron absorbing characteristics. In addition, in order to maintain material properties for long-term use this boronated material is not strengthened by aging treatment. This paper summarizes an outline of the boronated aluminum alloy for basket assemblies by powder metallurgy. (author)

Heat transfer in a fuel pin shipping container

International Nuclear Information System (INIS)

Ingham, J.G.

1980-01-01

Maximum cladding temperatures occur when the IDENT 1578 fuel pin shipping container is installed in the T-3 Cask. The maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 19-pin basket reaches 400 watts. Since 45% of the energy which is generated in the fuel escapes the 19-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 400/.55 = 727 watts. Similarly, the maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 40-pin basket reaches 465 watts. Since 33% of the energy which is generated in the fuel escapes the 40-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 465/.66 = 704 watts. The IDENT 1578 fuel pin shipping container therefore meets its thermal design criteria. IDENT 1578 can handle fuel pins with a decay heat load of 600 watts while maintaining the maximum fuel pin cladding temperature below 800 0 F. The emissivities which were determined from the test results for the basket tubes and container are relatively low and correspond to new, shiny conditions. As the IDENT 1578 container is exposed to high temperatures for extended periods of time during the transportation of fuel pins, the emissivities will probably increase. This will result in reduced temperatures

Removable cruciform for ice condenser ice basket

International Nuclear Information System (INIS)

Scrabis, C.M.; Mazza, G.E.; Golick, L.R.; Pomaibo, P.

1987-01-01

A removable cruciform for use in an ice basket having a generally cylindrical sidewall defining a central, vertical axis of the ice basket and plural, generally annular retaining rings secured to the interior of the cylindrical sidewall of the ice basket at predetermined, spaced elevations throughout the axial height of the ice basket is described comprising: a pair of brackets, each comprising a central, base portion having parallel longitudinal edges and a pair of integral legs extending at corresponding angles relative to the base portion from the perspective parallel longitudinal edges thereof; a pair of support plate assemblies secured to and extending in parallel, spaced relationship from one of the pair of brackets; a pair of slide support plates secured to the other of the pair of brackets and extending therefrom in spaced, parallel relationship; and spring means received within the housing and engaging the base portions of the brackets and applying a resilient biasing force thereto for maintaining the spaced relationship thereof

Bounds for Asian basket options

Science.gov (United States)

Deelstra, Griselda; Diallo, Ibrahima; Vanmaele, Michèle

2008-09-01

In this paper we propose pricing bounds for European-style discrete arithmetic Asian basket options in a Black and Scholes framework. We start from methods used for basket options and Asian options. First, we use the general approach for deriving upper and lower bounds for stop-loss premia of sums of non-independent random variables as in Kaas et al. [Upper and lower bounds for sums of random variables, Insurance Math. Econom. 27 (2000) 151-168] or Dhaene et al. [The concept of comonotonicity in actuarial science and finance: theory, Insurance Math. Econom. 31(1) (2002) 3-33]. We generalize the methods in Deelstra et al. [Pricing of arithmetic basket options by conditioning, Insurance Math. Econom. 34 (2004) 55-57] and Vanmaele et al. [Bounds for the price of discrete sampled arithmetic Asian options, J. Comput. Appl. Math. 185(1) (2006) 51-90]. Afterwards we show how to derive an analytical closed-form expression for a lower bound in the non-comonotonic case. Finally, we derive upper bounds for Asian basket options by applying techniques as in Thompson [Fast narrow bounds on the value of Asian options, Working Paper, University of Cambridge, 1999] and Lord [Partially exact and bounded approximations for arithmetic Asian options, J. Comput. Finance 10 (2) (2006) 1-52]. Numerical results are included and on the basis of our numerical tests, we explain which method we recommend depending on moneyness and time-to-maturity.

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

Interim report spent nuclear fuel retrieval system fuel handling development testing

Energy Technology Data Exchange (ETDEWEB)

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

1997-06-01

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

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

Bait preference in basket trap fishing operation and heavy metal ...

African Journals Online (AJOL)

The bait preference of basket traps fishing operation and heavy metal contamination in the trap catches from Lagos Lagoon were carried out between January and June 2011. Sixty baskets traps were used for the fishing operation, twenty basket traps were baited each with soap, coconut and maize. Clibanarius africanus ...

Sterilization of single-use helical stone baskets: an experimental study

Directory of Open Access Journals (Sweden)

Cely Barreto da Silva

2011-03-01

Full Text Available Objectives: To experimentally evaluate the efficacy of a standard sterilization protocol employed during reuse of disposable helical stone baskets. Methods: Study performed on 20 helical stone baskets: 10 were used in the initial validation process, contaminated with Escherichia coli ATCC 25922 and imprinted on Müeller-Hinton media; 10 catheters were contaminated with Geobacillus stearothermophilus ATCC 7953, processed, inoculated in TSB and incubated in a water bath at a temperature of 55ºC. Bacterial growth was evaluated after 1, 3, 5 and 7 days. After sterilization, stone baskets were also opened and closed 40 times to check for functional problems. All plastic and basket parts were carefully checked for damages. Results: After the 72-hour incubation period, there was growth of E. coli ATCC 25922 in 100% of imprints. After the sterilization process and up to 7 days incubation period on a blood agar plate, there was no growth of G. stearothermophilus ATCC 7953 or any other bacteria. There were no functional problems or damage to baskets after the sterilization process. Conclusion: The ethylene oxide system is efficacious and safe for sterilization of disposable helical stone baskets. However, further clinical studies are required and should provide more safety information.

Plant-scale anodic dissolution of unirradiated IFR fuel pins

International Nuclear Information System (INIS)

Gay, E.C.; Tomczuk, Z.; Miller, W.E.

1993-01-01

This report discusses anodic dissolution which is a major operation in the pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor (IFR), an advanced reactor design developed at Argonne National Laboratory. This process involves electrorefining the heavy metals (uranium and plutonium) from chopped, steel-clad fuel segments. The heavy metals are electrotransported from anodic dissolution baskets to solid and liquid cathodes in a molten salt electrolyte (LiCl-KCI) at 500 degrees C. Uranium is recovered on a solid cathode mandrel, while a uranium-plutonium mixture is recovered in a liquid cadmium cathode. The anode configuration consists of four baskets mounted on an anode shaft. These baskets provide parallel circuits in the electrolyte and salt flow through the chopped fuelbed as the baskets are rotated. The baskets for the engineering-scale tests were sized to contain up to 2.5 kg of heavy metal. Anodic dissolution of 10 kg batches of chopped, steel-clad simulated tuel (U-10% Zr and U-Zr-Fs alloy) was demonstrated

Implied Stopping Rules for American Basket Options from Markovian Projection

KAUST Repository

Bayer, Christian; Hä ppö lä , Juho; Tempone, Raul

2017-01-01

This work addresses the problem of pricing American basket options in a multivariate setting, which includes among others, the Bachelier and the Black-Scholes models. In high dimensions, nonlinear partial differential equation methods for solving the problem become prohibitively costly due to the curse of dimensionality. Instead, this work proposes to use a stopping rule that depends on the dynamics of a low-dimensional Markovian projection of the given basket of assets. It is shown that the ability to approximate the original value function by a lower-dimensional approximation is a feature of the dynamics of the system and is unaffected by the path-dependent nature of the American basket option. Assuming that we know the density of the forward process and using the Laplace approximation, we first efficiently evaluate the diffusion coefficient corresponding to the low-dimensional Markovian projection of the basket. Then, we approximate the optimal early-exercise boundary of the option by solving a Hamilton-Jacobi-Bellman partial differential equation in the projected, low-dimensional space. The resulting near-optimal early-exercise boundary is used to produce an exercise strategy for the high-dimensional option, thereby providing a lower bound for the price of the American basket option. A corresponding upper bound is also provided. These bounds allow to assess the accuracy of the proposed pricing method. Indeed, our approximate early-exercise strategy provides a straightforward lower bound for the American basket option price. Following a duality argument due to Rogers, we derive a corresponding upper bound solving only the low-dimensional optimal control problem. Numerically, we show the feasibility of the method using baskets with dimensions up to fifty. In these examples, the resulting option price relative errors are only of the order of few percent.

Implied Stopping Rules for American Basket Options from Markovian Projection

KAUST Repository

Bayer, Christian

2017-05-01

This work addresses the problem of pricing American basket options in a multivariate setting, which includes among others, the Bachelier and the Black-Scholes models. In high dimensions, nonlinear partial differential equation methods for solving the problem become prohibitively costly due to the curse of dimensionality. Instead, this work proposes to use a stopping rule that depends on the dynamics of a low-dimensional Markovian projection of the given basket of assets. It is shown that the ability to approximate the original value function by a lower-dimensional approximation is a feature of the dynamics of the system and is unaffected by the path-dependent nature of the American basket option. Assuming that we know the density of the forward process and using the Laplace approximation, we first efficiently evaluate the diffusion coefficient corresponding to the low-dimensional Markovian projection of the basket. Then, we approximate the optimal early-exercise boundary of the option by solving a Hamilton-Jacobi-Bellman partial differential equation in the projected, low-dimensional space. The resulting near-optimal early-exercise boundary is used to produce an exercise strategy for the high-dimensional option, thereby providing a lower bound for the price of the American basket option. A corresponding upper bound is also provided. These bounds allow to assess the accuracy of the proposed pricing method. Indeed, our approximate early-exercise strategy provides a straightforward lower bound for the American basket option price. Following a duality argument due to Rogers, we derive a corresponding upper bound solving only the low-dimensional optimal control problem. Numerically, we show the feasibility of the method using baskets with dimensions up to fifty. In these examples, the resulting option price relative errors are only of the order of few percent.

CFD Simulation of Heat and Fluid Flow for Spent Fuel in a Dry Storage

International Nuclear Information System (INIS)

In, Wangkee; Kwack, Youngkyun; Kook, Donghak; Koo, Yanghyun

2014-01-01

single basket by using the entrance boundary condition from the first cask analysis. The CFD results can be summarized as follows. An upward fluid (helium) flow was induced by decay heat inside the basket and a downward flow in the peripheral region of the cask by cooling. The helium velocity was accelerated to approximately 0.32 m/sec at the basket exit due to the buoyance effect being more significant in the outer basket. The helium temperature rise was predicted to be 200 .deg. C in the center basket and 90 .deg. C in the outer basket, respectively. The helium and fuel temperature in the central region of the fuel assembly is higher than the temperature in the outer region. The maximum fuel temperature in the center basket is estimated to be 330 .deg. C. The radiation heat transfer should be included in the future CFD analysis for the best estimate of the spent fuel temperature

Clostridium difficile in retail baskets, trolleys, conveyor belts, and plastic bags in Saudi Arabia.

Science.gov (United States)

Alqumber, Mohammed A

2014-10-01

To determine Clostridium difficile (C. difficile) prevalence on retail surfaces and shoppers plastic bags. From 20 June to 10 August 2011, in a cross-sectional epidemiological study, 17 supermarkets from 2 cities, Albaha and Altaif, Saudi Arabia were sampled. A total of 800 samples, which comprised 200 samples per surveyed surface, were studied. These included baskets, trolleys, conveyer belts, and outgoing shoppers' plastic bags. Clostridium difficile strains were isolated. The isolates were characterized using ribotyping and  polymerase chain reaction for the detection of toxin A (tcdA), toxin B (tcdB), binary toxin (cdtB), and toxin C (tcdC) genes. Susceptibility to antibiotics was determined on a Muller-Hinton agar with 5% sheep blood agar using E-tests. Overall, the C. difficile prevalence on sampled surfaces was 0.75%. The highest prevalence was found on retail baskets and trolleys, followed by plastic bags. A total of 5 different ribotypes were identified. Alterations in tcdC were detected in ribotype 027 and BT1. All the identified isolates were susceptible to vancomycin, but resistant to levofloxacin. In this study, C. difficile was present at a rate of 0.75% on supermarket surfaces. Spore disinfection of implicated surfaces may be necessary to control any community-acquired infections caused by this pathogen. 

Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

International Nuclear Information System (INIS)

Broadhead, B.L.

1991-08-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

The incorporation of boron in fissile transport packages for the transport and interim storage of irradiated light water reactor fuels

International Nuclear Information System (INIS)

Hunter, I.J.

1998-01-01

Boron is widely used in the nuclear industry for capturing neutrons and it is particularly useful in the criticality control of packages for the transport and interim storage of irradiated light water reactor fuels. Such fuels are typically located in an internal frame of stainless steel or aluminium, referred to as a basket, which locates the fuel assemblies in channels. Transnucleaire has designed and supplied more than 100 baskets of varying types during the past 30 years. Boron has been incorporated in many forms. Early designs of baskets used boron in specific zones to contribute to the control of criticality. Later developments in new materials dispersed boron throughout the basket and gave designers more options for the basic forms which make up the channels. New basket concepts have been developed by Transnucleaire to meet the changing market needs for transport and interim storage and boron continues to play an important role as an efficient thermal neutron absorber. (author)

A novel stone retrieval basket for more efficient lithotripsy procedures.

Science.gov (United States)

Salimi, N; Mahajan, A; Don, J; Schwartz, B

2009-01-01

This paper presents the development of an improved stone retrieval device that uses a newly designed cage of Nitinol wires encompassing a mesh basket made of a material that is laser resistant. Current methods to extract large stones involve imaging, using a laser to fragment the stones and then using existing cage-like baskets to trap the fragments individually and extracting them one at a time. These procedures are tedious, and may result in leaving some fragments behind that can reform causing the need for another procedure. The device presented in this paper will have a mesh-like sack which will consist of a laser resistant material of polytetrafluoroethylene (PTFE) enclosed within a newly designed Nitinol cage. Two alternate designs are provided for the cage in this paper. The handle of the device is revised to allow for a 3 Fr (1 mm) opening such that a laser's fiber optic cable can enter the device. Using this device a laser can be used to fragment the stone, and all the fragments are retained in the basket in both the design options. The basket can then be retracted allowing for the retrieval of all the fragments in one shot. The stone retrieval basket presented in this paper will significantly improve the efficiency and effectiveness of lithotripsy procedures for removal of large kidney and biliary tract stones.

Thermal analysis of cold vacuum drying of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Piepho, M.G.

1998-07-20

The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

An assessment of temperature history on concrete silo dry storage system for CANDU spent fuel

International Nuclear Information System (INIS)

Lee, Dong-Gyu; Sung, Nak-Hoon; Park, Jea-Ho; Chung, Sung-Hwan

2016-01-01

Highlights: • We performed thermal analysis to predict the temperature distribution in the concrete silo. • Thermal analysis of the concrete silo was based on CFD code. • Temperature distribution and history for storage period was presented. • Thermal analysis results and test results agreed well. • The correlations can predict the maximum fuel temperature over storage period. - Abstract: Concrete silo is a dry storage system for spent fuel generated from CANDU reactors. The silo is designed to remove passively the decay heat from spent fuel, as well as to secure the integrity of spent fuel during storage period. Dominant heat transfer mechanisms must be characterized and validated for the thermal analysis model of the silo, and the temperature history along storage period could be determined by using the validated thermal analysis model. Heat transfer characteristics on the interior and exterior of fuel basket in the silo were assessed to determine the temperature history of silo, which is necessary for evaluating the long-term degradation behavior of CANDU spent fuel stored in the silo. Also a methodology to evaluate the temperature history during dry storage period was proposed in this study. A CFD model of fuel basket including fuel bundles was suggested and temperature difference correlation between fuel bundles and silo’s internal member, as a function of decay heat of fuel basket considering natural convection and radiation heat transfer, was deduced. Temperature difference between silo’s internal cavity and ambient air was determined by using a concept of thermal resistance, which was validated by CFD analysis. Fuel temperature was expressed as a function of ambient temperature and decay heat of fuel basket in the correlation, and fuel temperature along storage period was determined. Therefore, it could be used to assess the degradation behavior of spent fuel by applying the degradation mechanism expressed as a function of spent fuel

Some topics in mathematical finance: Asian basket option pricing, Optimal investment strategies

OpenAIRE

Diallo, Ibrahima

2010-01-01

This thesis presents the main results of my research in the field of computational finance and portfolios optimization. We focus on pricing Asian basket options and portfolio problems in the presence of inflation with stochastic interest rates.In Chapter 2, we concentrate upon the derivation of bounds for European-style discrete arithmetic Asian basket options in a Black and Scholes framework.We start from methods used for basket options and Asian options. First, we use the general approach f...

2D and 3D thermal simulations for storage systems with internal natural convection for canistered spent fuel

International Nuclear Information System (INIS)

Yaksh, M.; Wang, C.

2004-01-01

In the US, the number of nuclear plants expected to implement on-site dry storage is increasing each year. As reactors burn advanced fuel assemblies to higher burnups, the dry storage systems will be required to accommodate higher heat loads. This is due to the increasing capacity of the systems and the need to store higher burnup fuel with reasonable cooling periods (i.e., five to six years). As the storage systems heat rejection design must be passive, natural convection is an efficient means for rejection of heat from the spent fuel to the surface of the canister boundary. The design presented in this paper is a canistered system that employs conduction, radiation and convection to reject heat from the canister, which is stored in a vertical concrete cask. The canister containing the spent fuel in this design is a right circular stainless steel vessel capable of storing 37 PWR fuel assemblies with a total canister heat load of 40 kW. Accompanying any design effort is the use of a numerical methodology that can accurately predict the peak-clad temperatures of the fuel and the structural components of the system. The main challenge to any analysis employing internal natural convection may be perceived as a practical limitation due to the size of the model. Since canisters are typically cylindrical, a two-dimensional model can be used to represent the canister. The fuel basket structure, which maintains the configuration of the spent fuel, is an array of square tubes, and is non-axisymmetric. Flow up through the fuel region in the basket encounters a complex cross section due to the fuel assembly rod array (up to 17 x 17). The flow region of the heated gas down the outside of the basket in the annulus between the canister shell and the basket assembly (downcomer) is also an irregular shaped area. To confirm that a two-dimensional (2D) modelling methodology is appropriate, a benchmark using results from a thermal test is required. The thermal test focuses on the

Spent nuclear fuel project product specification

International Nuclear Information System (INIS)

Pajunen, A.L.

1998-01-01

Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification

BASKET HASIL TANGKAPAN DAN KETERKAITANNYA DENGAN MUTU HASIL TANGKAPAN DAN SANITASI DI TPI PPN PALABUHANRATU

Directory of Open Access Journals (Sweden)

Anwar Bey Pane

2008-12-01

Full Text Available Fish basket has an important role in assisting efficacy of marketing and fish landing process at fishing port. Generally have been known the function of basket as place of fish conveyor. Basket that was used during the time in fishing port (FP and fish landing place (FLP, indication of apperance of uncleanness at port quay, fish auction place (FAP/TPI and it surrounding. The problem of sanitaion and also fish quality as ports at this period are very important for FP/FLP in Indonesia, especially for facing the era of globalization and free market. Including for FP such as PPN Palabuhanratu will be developed into ocean fishing port type; with one of function is to provide the fish for export purposes. This research was conducted at PPN Palabuhanratu at period of Sep-Nov 2007, the aims of this research are for knowing the effect of useful of basket to fish quality and the port quay sanitation, FAP and its surrounding. This research using study case method that is research the catch aspect in FP and fish basket aspect of catch landing and marketing activities and port quay and FAP. In PPN Palabuhanratu, the existing of fish basket are have so many type; plastic basket, bamboo basket, barrel, styrofoam box and container fibreglass (jolang. The use of its have affect to fish quality differently: assisting to maintain and degrade the quality of fish. Almost all type of fish basket have negative influence to sanitaion of port quay, FA and its surrounding, except box of styrofoam for layur. No one of the type of fish basket are able to provide only positive influence on fish quality and satation

Optimal Currency Basket Pegs for Developing and Emerging Economies

OpenAIRE

Joseph Daniels; Peter G. Toumanoff; Marc von der Ruhr

2001-01-01

The exchange rate arrangement represents an important policy choice for emerging and transitional economies as they strive to become stable and market-driven. A wide variety of arrangements have emerged, ranging from currency boards, basket-currency pegs and single-currency pegs to floating rates. Recently the IMF has recommended that, if the exchange value of a currency is to be pegged, it is better to peg to a basket of currencies rather than a single currency. Nonetheless, there has been l...

Storage, transportation and disposal system for used nuclear fuel assemblies

Science.gov (United States)

Scaglione, John M.; Wagner, John C.

2017-01-10

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

Mud-plastered granary-baskets at a Celtic Oppidum near Čarnok (Vojvodina, Serbia

Directory of Open Access Journals (Sweden)

Medović Aleksandar

2011-01-01

Full Text Available In the Celtic Oppidum near Čarnok (Bačka, Vojvodina, Serbia remains of four mud-plastered granarybaskets were discovered. They are all dated to the period Gomolava VI-B (first half of the 1st century B.C.. Hulled barley and broomcorn millet were stored in the granary-baskets. The mesh of the granary-baskets was composed of young branches of oak tree (Quercus, common English elm tree (Ulmus cf. campestris L., poplar / willow (Populus / Salix, purgin buckthorn (Rhamnus cf. cathartica L., wayfaring tree (Viburnum cf. lantana L., spindle tree (Euonymus cf. europaeus L., barberry (Berberis vulgaris L. and a Pomoidae-tree. At least four different tree taxa were used for basket making. Construction of two-thousand-year old mud-plastered granary-baskets from the Pannonian plain is very similar to that of traditional granaries in some present-day villages in Africa. Additionally, one non-plastered basket was also discovered at Oppidum. It was build out of maple (Acer, barberry, buckthorn (Rhamnus and branches of a pomaceous fruits tree (Pomoidae. .

Early effects of trimethyltin on the dentate gyrus basket cells: a morphological study

Energy Technology Data Exchange (ETDEWEB)

Chang, L.W.; Dyer, R.S.

1985-01-01

Electrophysiological evidence for reduction of recurrent inhibition in the dentate gyrus in animals exposed to trimethyltin (TMT) suggested alterations in the inhibitory neurons (basket cells) by TMT. The present study was designed to investigate the morphology of basket cells after TMT exposure. Long-Evans hooded rats were injected with TMT chloride in a dose of 6.0 mg/kg body weight (b.w.). Tissue samples from the dentate gyri were examined by both light and electron microscopy at 24 and 72 h after TMT exposure. Except for isolated basket cell damage at 72 h, no remarkable pathological changes were observed with light microscopy. Consistent with previous data, electron microscopy revealed that the basket cells of the dentate gyrus are large neurons situated just below the granule cell layer with characteristic large, infolded nuclei and intranuclear filamentous rods. Increased cytoplasmic density and degenerative changes of the Golgi complex were evident in the basket cells as early as 24 h after TMT exposure. By 72 h, neuronal vacuolation, accumulation of lysosomes, and occasional neuronal necrosis were observed. No significant pathological changes were found among the granule cells at this time. This report provides the first morphological evidence for early damage to the basket cells by TMT, which may account for the reduction of recurrent inhibition and hyperexcitability among the granule cells reported previously.

Vulnerability of Bread-Baskets to Weather Shocks

Science.gov (United States)

Gerber, J. S.; Ray, D. K.; West, P. C.; Foley, J. A.

2013-12-01

Many analyses of food security consider broad trends in food supply (crop production, crop use) and demand (changing diets, population growth.) However, if past shocks to the food system due to weather events (i.e. droughts) were to repeat themselves today, the resulting famines could be far more serious due to increased concentration of grain production in vulnerable bread-baskets, and decreased resilience of global and regional food systems (i.e. lower stocks, dependence on fewer crops). The present research project takes advantage of high-resolution historical weather datasets to assess probabilities of historically observed droughts repeating themselves in one or more of today's bread-basket regions. Using recently developed relationships between weather and crop yield, we consider the likelihood of region-wide crop failures under current conditions, and also under various climate scenarios.

Functional characteristics of parvalbumin- and cholecystokinin-expressing basket cells.

Science.gov (United States)

Bartos, Marlene; Elgueta, Claudio

2012-02-15

Cortical neuronal network operations depend critically on the recruitment of GABAergic interneurons and the properties of their inhibitory output signals. Recent evidence indicates a marked difference in the signalling properties of two major types of perisomatic inhibitory interneurons, the parvalbumin- and the cholecystokinin-containing basket cells. Parvalbumin-expressing basket cells are rapidly recruited by excitatory synaptic inputs, generate high-frequency trains of action potentials, discharge single action potentials phase-locked to fast network oscillations and provide fast, stable and timed inhibitory output onto their target cells. In contrast, cholecystokinin-containing basket cells are recruited in a less reliable manner, discharge at moderate frequencies with single action potentials weakly coupled to the phases of fast network oscillations and generate an asynchronous, fluctuating and less timed inhibitory output. These signalling modes are based on cell type-dependent differences in the functional and plastic properties of excitatory input synapses, integrative qualities and in the kinetics and dynamics of inhibitory output synapses. Thus, the two perisomatic inhibitory interneuron types operate with different speed and precision and may therefore contribute differently to the operations of neuronal networks.

Investigation of a Shock Absorber for Safeguard of Fuel Assemblies Failure

International Nuclear Information System (INIS)

Karalevicius, Renatas; Dundulis, Gintautas; Rimkevicius, Sigitas; Uspuras, Eugenijus

2006-01-01

The Ignalina NPP has two reactors. The Unit 1 was shut down, therefore the special equipment was designed for transportation of the fuel from Unit 1 to Unit 2. The fuel-loaded basket can drop during transportation. The special shock absorber was designed in order to avoid failure of fuel assemblies during transportation. In case of drop of fuel loaded basket, the failure of fuel assemblies can occur. This shock absorber was studied by scaled experiments at Lithuanian Energy Institute. Static and dynamic investigations of shock absorber are presented in this paper, including dependency of axial force versus axial compression. The finite element codes BRIGADE/Plus and ABAQUS/Explicit were used for analysis. Static simulation was used to optimize the dimensions of shock absorber. Dynamic analysis shows that shock absorber is capable to withstand the dynamic load for successful force suppression function in case of an accident. (authors)

Impact of a student's basket on work organization in schools

OpenAIRE

Zaleckienė, Lina; Šaparnis, Gintaras

2006-01-01

Support of the education system and its use of resources are being reformed in order to help the system to better adapt to free market and ensure its higher quality and better accessibility to it. Formation of a student's basket is assessed positively but head teachers do not consider it as a factor of changes in education so it is more often generalized as a possibility to plan and use finances and resources more efficiently. The formation of a student's basket is linked with in-service trai...

Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

International Nuclear Information System (INIS)

Tadano, Tomoaki

2016-01-01

The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

Evaluation of Metal-Fueled Surface Reactor Concepts

International Nuclear Information System (INIS)

Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

2007-01-01

Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

Primary Care Providers' Opening of Time-Sensitive Alerts Sent to Commercial Electronic Health Record InBaskets.

Science.gov (United States)

Cutrona, Sarah L; Fouayzi, Hassan; Burns, Laura; Sadasivam, Rajani S; Mazor, Kathleen M; Gurwitz, Jerry H; Garber, Lawrence; Sundaresan, Devi; Houston, Thomas K; Field, Terry S

2017-11-01

Time-sensitive alerts are among the many types of clinical notifications delivered to physicians' secure InBaskets within commercial electronic health records (EHRs). A delayed alert review can impact patient safety and compromise care. To characterize factors associated with opening of non-interruptive time-sensitive alerts delivered into primary care provider (PCP) InBaskets. We analyzed data for 799 automated alerts. Alerts highlighted actionable medication concerns for older patients post-hospital discharge (2010-2011). These were study-generated alerts sent 3 days post-discharge to InBaskets for 75 PCPs across a multisite healthcare system, and represent a subset of all urgent InBasket notifications. Using EHR access and audit logs to track alert opening, we performed bivariate and multivariate analyses calculating associations between patient characteristics, provider characteristics, contextual factors at the time of alert delivery (number of InBasket notifications, weekday), and alert opening within 24 h. At the time of alert delivery, the PCPs had a median of 69 InBasket notifications and had received a median of 379.8 notifications (IQR 295.0, 492.0) over the prior 7 days. Of the 799 alerts, 47.1% were opened within 24 h. Patients with longer hospital stays (>4 days) were marginally more likely to have alerts opened (OR 1.48 [95% CI 1.00-2.19]). Alerts delivered to PCPs whose InBaskets had a higher number of notifications at the time of alert delivery were significantly less likely to be opened within 24 h (top quartile >157 notifications: OR 0.34 [95% CI 0.18-0.61]; reference bottom quartile ≤42). Alerts delivered on Saturdays were also less likely to be opened within 24 h (OR 0.18 [CI 0.08-0.39]). The number of total InBasket notifications and weekend delivery may impact the opening of time-sensitive EHR alerts. Further study is needed to support safe and effective approaches to care team management of InBasket notifications.

Basket Option Pricing Using GP-GPU Hardware Acceleration

KAUST Repository

Douglas, Craig C.; Lee, Hyoseop

2010-01-01

We introduce a basket option pricing problem arisen in financial mathematics. We discretized the problem based on the alternating direction implicit (ADI) method and parallel cyclic reduction is applied to solve the set of tridiagonal matrices

Typewriting: In-Basket Exercises for Production Typewriting

Science.gov (United States)

Stamps, Charles A.

1977-01-01

Describes the use of in-basket exercises, a nontraditional approach to teaching production typewriting, which (according to the author) enables students to increase their production typewriting speed, develop the ability to rely on their own decisions and interpretations of written instructions, learn the necessity for organization, and compose…

Surface area considerations for corroding N reactor fuel

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Pitner, A.L.

1996-06-01

The N Reactor fuel is corroding at sites where the Zircaloy cladding was damaged when the fuel was discharged from the reactor. Corroding areas are clearly visible on the fuel stored in open cans in the K East Basin. There is a need to estimate the area of the corroding uranium to analyze aspects of fuel behavior as it is transitioned. from current wet storage to dry storage. In this report, the factors that contribute to open-quotes trueclose quotes surface area are analyzed in terms of what is currently known about the N Reactor fuel. Using observations from a visual examinations of the fuel in the K East wet storage facility, a value for the corroding geometric area is estimated. Based on observations of corroding uranium and surface roughness values for other metals, a surface roughness factor is also estimated and applied to the corroding K East fuel to provide an estimated open-quotes trueclose quotes surface area. While the estimated area may be modified as additional data become available from fuel characterization studies, the estimate provides a basis to assess effects of exposed uranium metal surfaces on fuel behavior in operations involved in transitioning from wet to dry storage, during shipment and staging, conditioning, and dry interim storage

Pyramid-like basket cells in the granular layer of the dentate gyrus in the rat.

Science.gov (United States)

Seress, L

1978-01-01

Basket cells of the dentate gyrus were identified using Nissl (cresyl violet) staining. It has been found that the ratio between basket and granule cells is 1:150--210. Only a few glial cells, mainly astroglia, were found in the granular layer of the dentate gyrus. In accordance with earlier data it was found that the granule cells and glial cells originate mainly postnatally, but the basket cells, like the pyramidal cells of the hippocampus, originate prenatally. Images Fig. 1 Fig. 2 Fig. 3 Fig. 4 Fig. 5 Fig. 6 PMID:701192

PENGARUH MODEL LATIHAN DAN KOORDINASI TERHADAP KETERAMPILAN SISWI EKSTRAKURIKULER BOLA BASKET SMPN I BANTUL

Directory of Open Access Journals (Sweden)

Nurhidayah Nurhidayah

2015-04-01

Full Text Available Penelitian ini bertujuan untuk mengetahui (1 perbedaan pengaruh model latihan guided discovery style dan model latihan command style terhadap keterampilan bola basket siswi, (2 perbeda-an pengaruh siswi yang memiliki koordinasi tinggi dan koordinasi rendah terhadap keterampilan bola basket, dan (3 interaksi antara model latihan dengan koordinasi terhadap keterampilan bola basket siswi. Penelitian ini merupakan penelitian eksperimen. Populasi penelitian adalah seluruh siswi ekstrakurikuler bola basket yang berjumlah 43 orang. Sampel berjumlah 24 siswi ditentukan dengan teknik purposive sampling. Intrumen tes yang digunakan adalah tes Wall Bounce Pass untuk mengukur koordinasi mata-tangan dan tes Sekolah Tinggi Olahraga (STO untuk mengukur keterampilan bola basket siswi. Teknik analisis data yang menggunakan ANAVA. Hasil penelitian adalah sebagai berikut: (1 Tidak terdapat perbedaan pengaruh antara model latihan guided discovery style dan model latihan command style terhadap keterampilan bola basket siswi. (2 Terdapat perbedaan pengaruh siswi yang memiliki koordinasi tinggi dan rendah terhadap keterampilan bola basket. (3 Tidak terdapat interaksi yang signifikan antara model latihan dan koordinasi terhadap keterampilan bola basket siswi.   THE EFFECT OF TRAINING MODEL AND COORDINATION ON THE BASKETBALL SKILLS OF EXTRACURRICULAR STUDENTS OF JUNIOR HIGH SCHOOL 1 BANTUL   Abstract This study aims to investigate: (1 the different effects of the guided discovery style model training and command style model training on basketball skills of students, (2 the different effects on basketball skills of students who have high and low coordination, and (3 the interaction effects between training models of guided discovery style and command style with coordination (high and low on the basketball skills of students.This research was an experiment. The population comprised 43 basketball extracurricular students. A sample of 24 students was established using the

Handling and prevention of complications in stone basketing

NARCIS (Netherlands)

de La Rosette, Jean J. M. C. H.; Skrekas, Thomas; Segura, Joseph W.

2006-01-01

OBJECTIVE: To review the incidence, risk factors, and treatment options of intraoperative and postoperative complications of stone basketing in urology with emphasis on certain principles that must be upheld to prevent those complications. METHODS: A literature search was performed using the MEDLINE

Development of a Computerized In-Basket Exercise for the Classroom: A Sales Management Example

Science.gov (United States)

Pearson, Michael M.; Barnes, John W.; Onken, Marina H.

2006-01-01

This article follows the development of a sales management in-basket exercise for use in the classroom. The authors have computerized the exercise and added features to allow for additional and more quantitative input from the students. The exercise has evolved and been tested in numerous classroom situations. The computerized in-basket exercise…

Non-destructive assay of leached hulls in a nuclear fuel reprocessing plant

International Nuclear Information System (INIS)

Hofstetter, K.J.; Henderson, B.C.; Gray, J.H.; Huff, G.A.

1978-01-01

The hull monitor at the Barnwell Nuclear Fuels Plant (BNFP) will be a remotely controlled, fully automated system designed to quantitatively assay leached hulls for undissolved U and Pu. The hull monitor will assay the hulls from one metric ton of fuel per dissolver basket with the design goal of detecting 0.1% undissolved fuel and yet remain within the framework of the BNFP materials flow of five hull baskets per day. The non-destructive assay will be accomplished using a computer-based gamma-ray pulse height analysis system employing a 5 x 5 inch NaI(Tl) scintillation detector. The intense radiations from the fission product isotopes and the activation product isotopes produced in the reactor prevent direct assay of the undissolved fuel left in the hulls. The measurement will be made indirectly by demonstrating a correlation between the amount of 144 Ce undissolved and the remaining U. The isotope 144 Ce is a direct fission product with high cumulative yield. The daughter isotope 144 Pr has a gamma ray at 2.18 MeV well above other predominant radiations in the spectrum from the major interferences 60 Co, 58 Co, 95 Zr( 95 Nb), 137 Cs and 106 Ru( 106 Rh). Segmented scanning operation of the hull monitor is accomplished by rotation and vertical transversal of the hulls container past the detector station. Proper collimation and absorbers are required to maximize the 144 Ce( 144 Pr) to background ratio. A basket indexer is provided which monitors the scanning rate and ensures repositioning. The leached hull monitor system will be interfaced to a computer-based multichannel analyzer for ease of operation and data handling. A calibration basket has been fabricated to accomodate radioactive sources and inactive Zircaloy hulls

Design package lazy susan for the fuel retrieval system

International Nuclear Information System (INIS)

TEDESCHI, D.J.

1999-01-01

This is a design package that contains the details for a Lazy Susan style small tool for the Fuel Retrieval System. The Lazy Susan tool is used to help rotate an MCO Fuel Basket when loading it. This document contains requirements, development design information, tests and test reports that pertain to the production of Lazy Susan small tool

Advancing the design of a dynamic petro-dollar currency basket

International Nuclear Information System (INIS)

Elbeck, Matt

2010-01-01

This study offers Organization of Petroleum Exporting Countries (OPEC) member nations a crude oil pricing currency basket based on currency liquidity, in contrast with prior emphasis on OPEC trading patterns. Motivating the search for an alternative US dollar pricing of crude oil is the significant and inverse relationship (r=-0.82, p<0.01) between the US dollar major currencies index and crude oil price over the period January 1999-March 2009. A dynamically weighted petro-dollar currency basket is proposed based on the five currency claims (US dollar, Euro, British pound, Japanese yen and Swiss franc) and their varying proportions of foreign exchange reserves held by central banks. The major currencies US dollar index is compared against the proposed petro-dollar index to reveal an average US$8.1 billion annual gain in favor of the petro-dollar currency basket, offering OPEC members a revenue stream of diversified and highly liquid currencies to transition away from complete dependence on the US dollar crude oil pricing. The proposed petro-dollar crude oil pricing schema offers OPEC a crude oil price dynamically denominated in currencies reflecting the global use and importance of crude oil. This paper concludes with implementation issues facing a move toward the dynamically weighted petro-dollar crude oil pricing schema. (author)

SNF fuel retrieval sub project safety analysis document

International Nuclear Information System (INIS)

BERGMANN, D.W.

1999-01-01

This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed

SNF fuel retrieval sub project safety analysis document

Energy Technology Data Exchange (ETDEWEB)

BERGMANN, D.W.

1999-02-24

This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

Use of Linear Programming to Develop Cost-Minimized Nutritionally Adequate Health Promoting Food Baskets.

Science.gov (United States)

Parlesak, Alexandr; Tetens, Inge; Dejgård Jensen, Jørgen; Smed, Sinne; Gabrijelčič Blenkuš, Mojca; Rayner, Mike; Darmon, Nicole; Robertson, Aileen

2016-01-01

Food-Based Dietary Guidelines (FBDGs) are developed to promote healthier eating patterns, but increasing food prices may make healthy eating less affordable. The aim of this study was to design a range of cost-minimized nutritionally adequate health-promoting food baskets (FBs) that help prevent both micronutrient inadequacy and diet-related non-communicable diseases at lowest cost. Average prices for 312 foods were collected within the Greater Copenhagen area. The cost and nutrient content of five different cost-minimized FBs for a family of four were calculated per day using linear programming. The FBs were defined using five different constraints: cultural acceptability (CA), or dietary guidelines (DG), or nutrient recommendations (N), or cultural acceptability and nutrient recommendations (CAN), or dietary guidelines and nutrient recommendations (DGN). The variety and number of foods in each of the resulting five baskets was increased through limiting the relative share of individual foods. The one-day version of N contained only 12 foods at the minimum cost of DKK 27 (€ 3.6). The CA, DG, and DGN were about twice of this and the CAN cost ~DKK 81 (€ 10.8). The baskets with the greater variety of foods contained from 70 (CAN) to 134 (DGN) foods and cost between DKK 60 (€ 8.1, N) and DKK 125 (€ 16.8, DGN). Ensuring that the food baskets cover both dietary guidelines and nutrient recommendations doubled the cost while cultural acceptability (CAN) tripled it. Use of linear programming facilitates the generation of low-cost food baskets that are nutritionally adequate, health promoting, and culturally acceptable.

A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

International Nuclear Information System (INIS)

Kim, Chang Hyun

1997-02-01

A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

Disposal criticality analysis for aluminum-based DOE fuels

International Nuclear Information System (INIS)

Davis, J.W.; Gottlieb, P.

1997-11-01

This paper describes the disposal criticality analysis for canisters containing aluminum-based Department of Energy fuels from research reactors. Different canisters were designed for disposal of highly enriched uranium (HEU) and medium enriched uranium (MEU) fuel. In addition to the standard criticality concerns in storage and transportation, such as flooding, the disposal criticality analysis must consider the degradation of the fuel and components within the waste package. Massachusetts Institute of Technology (MIT) U-Al fuel with 93.5% enriched uranium and Oak Ridge Research Reactor (ORR) U-Si-Al fuel with 21% enriched uranium are representative of the HEU and MEU fuel inventories, respectively. Conceptual canister designs with 64 MIT assemblies (16/layer, 4 layers) or 40 ORR assemblies (10/layer, 4 layers) were developed for these fuel types. Borated stainless steel plates were incorporated into a stainless steel internal basket structure within a 439 mm OD, 15 mm thick XM-19 canister shell. The Codisposal waste package contains 5 HLW canisters (represented by 5 Defense Waste Processing Facility canisters from the Savannah River Site) with the fuel canister placed in the center. It is concluded that without the presence of a fairly insoluble neutron absorber, the long-term action of infiltrating water can lead to a small, but significant, probability of criticality for both the HEU and MEU fuels. The use of 1.5kg of Gd distributed throughout the MIT fuel and the use of carbon steels for the structural basket or 1.1 kg of Gd distributed in the ORR fuel will reduce the probability of criticality to virtually zero for both fuels

In vitro head-to-head comparison of the durability, versatility and efficacy of the NGage and novel Dakota stone retrieval baskets.

Science.gov (United States)

Bechis, Seth K; Abbott, Joel E; Sur, Roger L

2017-12-01

To compare head to head two end-engaging nitinol stone retrieval devices available to urologists, in terms of durability, versatility and efficacy. For durability testing, 30 NGage and Dakota baskets were cycled 20 times between grasping and releasing synthetic stone models and evaluated for damage or device failure. For versatility and efficacy testing, baskets were assessed in their ability to capture and release stone models from 1 to 11 mm. Each stone was raised above the capture site and the basket was opened to passively release the stone. If the stone did not release, the basket handle was shaken and the OpenSure feature employed if needed. Manual release was used as a last resort. Durability-the Cook NGage demonstrated a statistically significant increased rate of visible device breakdown (P=0.0046) in 8 of 30 (26.7%) devices vs. 0 of 30 Dakota devices, with mean damage at 13.5 cycles. Versatility and efficacy-both 8 mm baskets successfully captured stones from 1-8 mm. The Dakota more effectively released 7-8 mm stones (PDakota. For 11 mm baskets, the Dakota released all stones up to 10 mm with simple opening, while the NGage released 10 of 15 (67%) of 9 mm stones and 1 of 15 (7%) of 10 mm stones by simple opening. For 11 mm stones, the Dakota captured 100% whereas NGage could not capture any. Both baskets showed similar durability characteristics. The Dakota basket more effectively captured and released stones over 7 mm, as compared to the NGage basket. The OpenSure aspect conferred an advantage in handling and release of larger stones. These in vitro results demonstrate potential versatility, durability and efficacy of the Dakota basket.

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

An integrated fiber and stone basket device for use in Thulium fiber laser lithotripsy

Science.gov (United States)

Wilson, Christopher R.; Hutchens, Thomas C.; Hardy, Luke A.; Irby, Pierce B.; Fried, Nathaniel M.

2014-03-01

The Thulium fiber laser (TFL) is being explored as an alternative laser lithotripter to the Holmium:YAG laser. The TFL's superior near-single mode beam profile enables higher power transmission through smaller fibers with reduced proximal fiber tip damage. Recent studies have also reported that attaching hollow steel tubing to the distal fiber tip decreases fiber degradation and burn-back without compromising stone ablation rates. However, significant stone retropulsion was observed, which increased with pulse rate. In this study, the hollow steel tip fiber design was integrated with a stone basket to minimize stone retropulsion during ablation. A device was constructed consisting of a 100-μm-core, 140-μm-OD silica fiber outfitted with 5-mm-long stainless steel tubing at the distal tip, and integrated with a 1.3-Fr (0.433-mm-OD) disposable nitinol wire basket, to form an overall 1.9-Fr (0.633-mm- OD) integrated device. This compact design may provide several potential advantages including increased flexibility, higher saline irrigation rates through the ureteroscope working channel, and reduced fiber tip degradation compared to separate fiber and stone basket manipulation. TFL pulse energy of 31.5 mJ with 500 μs pulse duration and pulse rate of 500 Hz was delivered through the integrated fiber/basket device in contact with human uric acid stones, ex vivo. TFL stone ablation rates measured 1.5 +/- 0.2 mg/s, comparable to 1.7 +/- 0.3 mg/s (P > 0.05) using standard bare fiber tips separately with a stone basket. With further development, this device may be useful for minimizing stone retropulsion, thus enabling more efficient TFL lithotripsy at higher pulse rates.

Spatial variability in branchial basket meristics and morphology of ...

African Journals Online (AJOL)

We examined spatial variability in meristic and morphological characteristics of the branchial basket of sardine Sardinops sagax collected from four geographical regions around the southern African coast, namely Namibia and the South African west, south and east coasts. Our analysis tested the hypothesis of three putative ...

Pricing for a basket of LCDS under fuzzy environments.

Science.gov (United States)

Wu, Liang; Liu, Jie-Fang; Wang, Jun-Tao; Zhuang, Ya-Ming

2016-01-01

This paper looks at both the prepayment risks of housing mortgage loan credit default swaps (LCDS) as well as the fuzziness and hesitation of investors as regards prepayments by borrowers. It further discusses the first default pricing of a basket of LCDS in a fuzzy environment by using stochastic analysis and triangular intuition-based fuzzy set theory. Through the 'fuzzification' of the sensitivity coefficient in the prepayment intensity, this paper describes the dynamic features of mortgage housing values using the One-factor copula function and concludes with a formula for 'fuzzy' pricing the first default of a basket of LCDS. Using analog simulation to analyze the sensitivity of hesitation, we derive a model that considers what the LCDS fair premium is in a fuzzy environment, including a pure random environment. In addition, the model also shows that a suitable pricing range will give investors more flexible choices and make the predictions of the model closer to real market values.

Smoothing the payoff for efficient computation of Basket option prices

KAUST Repository

Bayer, Christian; Siebenmorgen, Markus; Tempone, Raul

2017-01-01

We consider the problem of pricing basket options in a multivariate Black–Scholes or Variance-Gamma model. From a numerical point of view, pricing such options corresponds to moderate and high-dimensional numerical integration problems with non

Economics of water basin storage of spent light water reactor fuel

International Nuclear Information System (INIS)

Driggers, F.E.

1978-01-01

As part of the International Spent Fuel Storage program, a preliminary Venture Guidance Assessment of the cost was made. The escalated cost of a reference facility with a capacity to receive 2000 MT/y of spent LWR fuel and to store 5000 MT in water-filled pools was converted to $180 million in 1978 dollars for a stand-alone facility. It was estimated that the receiving rate could be increased to 3000 MT/y for an additional $15 million and that increments could be added to the storage capacity for $13 million per 1000 MT. If a receipt rate of more than 3000 MT/y is required, a new facility in another part of the country might be built to reduce total costs including transportation. Operating costs are determined by the number of people employed and by the costs of stainless steel baskets. An operating crew of 150 is required for the reference facility; the associated cost, including overhead and supplies, is $6 million. During an extended storage-only period, this cost is assumed to drop to $4 million. Fuel baskets are estimated to cost $6.20/kg of spent fuel averaged over a reactor mix of two-thirds PWRs and one-third BWRs. The nominal basket requirements of $10 million for the first year are capitalized. If the facility is financed by the government and a one-time fee is charged to recover all of the away-from-reactor (AFR) basin costs, the fee is about $60/kg of spent fuel plus any government surcharge to cover research and development, overhead, and additional contingencies. If the facility is financed by industry with an annual charge that includes a fixed charge on capital of 25%, the annual fee is about $16/kg-y. In calculating both fees, it is assumed that each storage position is occupied for ten years. 8 tables

Physiological properties of anatomically identified basket and bistratified cells in the CA1 area of the rat hippocampus in vitro.

Science.gov (United States)

Buhl, E H; Szilágyi, T; Halasy, K; Somogyi, P

1996-01-01

Basket and bistratified cells form two anatomically distinct classes of GABAergic local-circuit neurons in the CA1 region of the rat hippocampus. A physiological comparison was made of intracellularly recorded basket (n = 13) and bistratified neurons (n = 6), all of which had been anatomically defined by their efferent target profile (Halasy et al., 1996). Basket cells had an average resting membrane potential of -64.2 +/- 7.2 vs. -69.2 +/- 4.6 mV in bistratified cells. The latter had considerably higher mean input resistances (60.2 +/- 42.1 vs. 31.3 +/- 10.9 M Ohms) and longer membrane time constants (18.6 +/- 8.1 vs. 9.8 +/- 4.5 ms) than basket cells. Differences were also apparent in the duration of action potentials, those of basket cells being 364 +/- 77 and those of bistratified cells being 527 +/- 138 microseconds at half-amplitude. Action potentials were generally followed by prominent, fast after-hyperpolarizing potentials which in basket cells were 13.5 +/- 6.7 mV in amplitude vs. 10.5 +/- 5.1 in bistratified cells. The differences in membrane time constant, resting membrane potential, and action potential duration reached statistical significance (P D-aspartate (NMDA) receptors, whereas the remaining slow-rise EPSP could be abolished by an NMDA receptor antagonist. Increasing stimulation intensity elicited biphasic inhibitory postsynaptic potentials (IPSPs) in both basket and bistratified cells. In conclusion, basket and bistratified cells in the CA1 area show prominent differences in several of their membrane and firing properties. Both cell classes are activated by Schaffer collateral/commissural axons in a feedforward manner and receive inhibitory input from other, as yet unidentified, local-circuit neurons.

C A R A fuel element for Atucha nuclear power plants and development plan

International Nuclear Information System (INIS)

Brasnarof, D. O; Marino, A. C; Bianchi, D; Giorgis M A; Orlando, O; Munoz, C; Taboada, H; Florido, P. C

2006-01-01

This paper presents the current state and the development plan of the C A R A fuel element.Main activities were carried out towards to welding of the end plates of the C A R A fuel element by a new process, and the assembling and hanging of the C A R A fuel element in its Atucha configuration, by using an external basket [es

Hot Surface Ignition of A Composite Fuel Droplet

Directory of Open Access Journals (Sweden)

Glushkov Dmitrii O.

2015-01-01

Full Text Available The present study examines the characteristics of conductive heating (up to ignition temperature of a composite fuel droplet based on coal, liquid petroleum products, and water. In this paper, we have established the difference between heat transfer from a heat source to a fuel droplet in case of conductive (hot surface and convective (hot gas heat supply. The Leidenfrost effect influences on heat transfer characteristics significantly due to the gas gap between a composite fuel droplet and a hot surface.

Use of Linear Programming to Develop Cost-Minimized Nutritionally Adequate Health Promoting Food Baskets

DEFF Research Database (Denmark)

Parlesak, A.; Tetens, Inge; Dejgård Jensen, Jørgen

2016-01-01

Food-Based Dietary Guidelines (FBDGs) are developed to promote healthier eating patterns, but increasing food prices may make healthy eating less affordable. The aim of this study was to design a range of cost-minimized nutritionally adequate health-promoting food baskets (FBs) that help prevent...... both micronutrient inadequacy and diet-related non-communicable diseases at lowest cost. Average prices for 312 foods were collected within the Greater Copenhagen area. The cost and nutrient content of five different cost-minimized FBs for a family of four were calculated per day using linear...... in each of the resulting five baskets was increased through limiting the relative share of individual foods. The one-day version of N contained only 12 foods at the minimum cost of DKK 27 (€ 3.6). The CA, DG, and DGN were about twice of this and the CAN cost ~DKK 81 (€ 10.8). The baskets with the greater...

Use of Linear Programming to Develop Cost-Minimized Nutritionally Adequate Health Promoting Food Baskets

DEFF Research Database (Denmark)

Parlesak, A.; Tetens, Inge; Dejgård Jensen, Jørgen

2016-01-01

programming. The FBs were defined using five different constraints: cultural acceptability (CA), or dietary guidelines (DG), or nutrient recommendations (N), or cultural acceptability and nutrient recommendations (CAN), or dietary guidelines and nutrient recommendations (DGN). The variety and number of foods...... in each of the resulting five baskets was increased through limiting the relative share of individual foods. The one-day version of N contained only 12 foods at the minimum cost of DKK 27 (€ 3.6). The CA, DG, and DGN were about twice of this and the CAN cost ~DKK 81 (€ 10.8). The baskets with the greater...... variety of foods contained from 70 (CAN) to 134 (DGN) foods and cost between DKK 60 (€ 8.1, N) and DKK 125 (€ 16.8, DGN). Ensuring that the food baskets cover both dietary guidelines and nutrient recommendations doubled the cost while cultural acceptability (CAN) tripled it. Use of linear programming...

Potential for Fuel Ignition after K Basin Drainage (Fauske and Associates Report FAI/99-71 Rev.1)

International Nuclear Information System (INIS)

DUNCAN, D.R.

1999-01-01

The potential for N reactor fuel ignition after hypothetical K basin drainage is considered here for fuel configurations and boundary conditions specified by the Spent Nuclear Fuel Project (SNFP). Configurations include: (1) Scrap canisters (open K East canisters containing primarily fragmented fuel) partially covered by sludge (on the exterior); (2) IWTS (Integrated Water Treatment System) settlers filled with fine fuel particulate; (3) IWTS knock out pots filled with coarse fuel particulate; (4) Scrap (fragmented fuel) in stylized configurations residing on the process table, including hemispherical and cylindrical piles; and (5) Scrap in a scrap basket on the process table. Fuel mass, metal fraction, and surface area or ranges for these parameters are specified by the SNFP in each configuration. Fuel and container exteriors are specified to be dry after the hypothetical drainage event, except in the case of fine particulate in the settlers which physically must hold water. Credibility of the specified scenarios and input parameters is neither endorsed nor judged in this report. The purpose of the calculations is to determine thermal stability of fuel given the specified configurations, parameters, and boundary conditions

Pulmonary vein isolation using the Rhythmia mapping system: Verification of intracardiac signals using the Orion mini-basket catheter.

Science.gov (United States)

Anter, Elad; Tschabrunn, Cory M; Contreras-Valdes, Fernando M; Li, Jianqing; Josephson, Mark E

2015-09-01

During pulmonary vein isolation (PVI), a circular lasso catheter is positioned at the junction between the left atrium (LA) and the pulmonary vein (PV) to confirm PVI. The Rhythmia mapping system uses the Orion mini-basket catheter with 64 electrodes instead of the lasso catheter. However, its feasibility to determine PVI has not been studied. The purpose of this study was to compare signals between the mini-basket and lasso catheters at the LA-PV junction. In 12 patients undergoing PVI using Rhythmia, the mini-basket and lasso catheters were placed simultaneously at the LA-PV junction for baseline and post-PVI signal assessment. Pacing from both catheters was performed to examine the presence of exit block. At baseline, recordings of LA and PV potentials were concordant in all PVs. However, after PVI, concordance between the catheters was only 68%. Discordance in all cases resulted from loss of PV potentials on the lasso catheter with persistence of PV potentials on the mini-basket catheter. In 9 of 13 PVs (69%), these potentials represented true PV potentials that were exclusively recorded with the smaller and closely spaced mini-basket electrodes. In the other 4 PVs (31%), these potentials originated from neighboring structures and resulted in underestimation of PVI. The use of the mini-basket catheter alone is sufficient to determine PVI. While it improves recording of PV potentials after incomplete ablation, it is also associated with frequent recording of "PV-like" potentials originating from neighboring structures. In these cases, pacing maneuvers are helpful to determine PVI and avoid excessive ablation. Copyright © 2015 Heart Rhythm Society. Published by Elsevier Inc. All rights reserved.

MARKET BASKET ANALYSIS (MBA PADA SITUS WEB E-COMMERCE ZAKIYAH COLLECTION

Directory of Open Access Journals (Sweden)

Ari Muzakir

2016-11-01

Full Text Available E-commerce menghubungkan antara produsen dengan produsen, produsen dengan konsumen, konsumen dengan produsen, konsumen dengan konsumen. Untuk mengimplementasi e-commerce dalam mendukung bisnis organisasi perlu di perhatikan 5 komponen utama yaitu ; pengembangan produk, promosi, transaksi online, product delivery dan after sales support. Hal ini yang tengah diterapkan pada Zakiyah Collection. Zakiyah Collection bergerak dibidang penjualan aneka macam kain khas Palembang seperti songket, blongket,tanjung, dan lain sebagainya. Untuk melakukan analisis terhadap pangsa pasar yang ada agar dapat bersaing dengan toko online lainnya dilakukan dengan strategi pemasaran dengan menggunakan pendekatan market basket analysis (MBA. MBA merupakan salah satu teknik dari data mining yang digunakan untuk menentukan produk-produk manakah yang akan dibeli oleh pelanggan secara bersamaan dengan melakukan analisa terhadap daftar transaksi pelanggan. Dengan mengetahui produk-produk tersebut, maka sebuah sistem e-commerce dapat membuat maupun mengembangkan sebuah sistem customer profiles dan dapat menentukan layout katalog pelanggannya sendiri. Model pengembangan sistem yang dilakukan menggunakan prototype dimana pelanggan dan pengguna akan dilibatkan secara langsung dalam proses ini. Hasil akhir dalam penelitian ini adalah berupa analisis data transaksi menggunakan market basket analysis dengan dilakukan 4 kali kombinasi produk yang berdasarkan nilai support x confidence terbesar dengan hasil berupa angka-angka kemungkinan transasksi yang berkaitan dengan produk yang dijual. Jika dengan menggunakan 1 kali kombinasi, maka didapatkan blongket dengan nilai support sebesar 0.5625. Jika dilakukan 2 kali kombinasi diperoleh kombinasi blongket dan songket dengan nilai support 0.375. Kata kunci: e-commerce, market basket analysis, association rules.

Basket Option Pricing Using GP-GPU Hardware Acceleration

KAUST Repository

Douglas, Craig C.

2010-08-01

We introduce a basket option pricing problem arisen in financial mathematics. We discretized the problem based on the alternating direction implicit (ADI) method and parallel cyclic reduction is applied to solve the set of tridiagonal matrices generated by the ADI method. To reduce the computational time of the problem, a general purpose graphics processing units (GP-GPU) environment is considered. Numerical results confirm the convergence and efficiency of the proposed method. © 2010 IEEE.

Experimental observations to the electrical field for electrorefining of spent nuclear fuel in the Mark-IV electrorefiner

International Nuclear Information System (INIS)

Li, S. X.

1998-01-01

Experimental results from the pilot scale electrorefiner (Mark-IV ER) treating spent nuclear fuel are reported in this article. The electrorefining processes were carried out in a LiCl-KCl-UCl 3 electrolyte. It has been noted that spool of molten cadmium below the electrolyte plays an important role in the electrorefining operations. In addition, formations of electrical shorting path between anode baskets and the electrorefiner vessel were observed, which lessened the uranium dissolution process from anode baskets, however appeared to improve the morphology of cathode deposit. The FIDAP simulation code was used to calculate the electrical potential field distributions and the potential gradient near the cathode. The effect of the electrical shorting between anode baskets and electrorefiner vessel on the morphology of cathode products is discussed

Re-Learning the Traditional Art of Inuit Grass Basket-Making

Science.gov (United States)

Cowan, Cindy

2005-01-01

This paper describes an adult learning project to revitalise the traditional Inuit art of weaving grass baskets. The participants involved in the project, all older women who speak an indigenous first language (Inuktitut) and who have limited experience with formal education, largely on their own initiative, undertook the process of successfully…

Use of the Escape nitinol stone retrieval basket facilitates fragmentation and extraction of ureteral and renal calculi: a pilot study.

Science.gov (United States)

Kesler, Stuart S; Pierre, Sean A; Brison, Daniel I; Preminger, Glenn M; Munver, Ravi

2008-06-01

Advances in ureteroscope and stone basket design have catapulted ureteroscopy to the forefront of surgical stone management; however, persistent problems such as stone migration continue to challenge urologists. The Escape nitinol stone retrieval basket (Boston Scientific, Natick, MA) is a stone basket designed to capture calculi and facilitate simultaneous laser lithotripsy in situ. We report our initial experience with the Escape basket for the management of urinary calculi and compare the use of this device with other methods of optimizing ureteroscopic stone management. A prospective evaluation of 23 patients undergoing ureteroscopic holmium:yttrium-aluminum-garnet laser lithotripsy of urinary calculi was performed at two institutions by two surgeons (R.M. and G.M.P). The Escape basket was used to prevent retrograde ureteral stone migration or to facilitate fragmentation and extraction of large renal calculi. Patient demographics and perioperative parameters were assessed. Twenty-three patients (16 men, 7 women), with a mean age of 55.5 years (range 33-74 yrs) were treated for renal (n = 9) or ureteral (n = 14) calculi. The mean stone diameter was 1.4 cm (range 0.4-2.5 cm), mean fragmentation time was 44.1 minutes (range 10-75 min), and mean energy used was 3.1 kJ (range 0.4-10.6 kJ). No complications were encountered. Eighty-seven percent (20/23) of patients were rendered completely stone free after ureteroscopic laser lithotripsy using the Escape basket. Of the three patients with residual calculi, one patient with a 2.5-cm renal calculus had residual fragments larger than 3 mm, and two patients with large renal calculi had residual fragments smaller than 3 mm. The Escape basket appears to be safe and effective in preventing stone migration and facilitating ureteroscopic laser lithotripsy and stone extraction.

The relationship of post-fire white ash cover to surface fuel consumption

Science.gov (United States)

Andrew T. Hudak; Roger D. Ottmar; Robert E. Vihnanek; Nolan W. Brewer; Alistair M. S. Smith; Penelope Morgan

2013-01-01

White ash results from the complete combustion of surface fuels, making it a logically simple retrospective indicator of surface fuel consumption. However, the strength of this relationship has been neither tested nor adequately demonstrated with field measurements. We measured surface fuel loads and cover fractions of white ash and four other surface materials (green...

Development of spent fuel dry storage technology

International Nuclear Information System (INIS)

Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

2000-01-01

The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

Bayesian techniques for surface fuel loading estimation

Science.gov (United States)

Kathy Gray; Robert Keane; Ryan Karpisz; Alyssa Pedersen; Rick Brown; Taylor Russell

2016-01-01

A study by Keane and Gray (2013) compared three sampling techniques for estimating surface fine woody fuels. Known amounts of fine woody fuel were distributed on a parking lot, and researchers estimated the loadings using different sampling techniques. An important result was that precise estimates of biomass required intensive sampling for both the planar intercept...

Importance of public participation in decision-making process in healthcare system illustarted with an example of the development of American and Polish scope of health benefit basket.

Science.gov (United States)

Kolasa, Katarzyna; Hermanowski, Tomasz; Borek, Ewa

2013-01-01

The process of the development of health benefit basket may serve as a good example of decision-making process in the healthcare system which is based on public participation. Comparative analysis of development and implementation of health benefit basket in Poland and the USA. On a basis of the literature review, following questions were studied, i.e.: What is the origin of health benefit basket development in the USA and Poland? What was the role of pubic opinion in determining the range of health benefit basket in both countries? What criteria were employed to determine the range of health benefit basket in both countries? What conclusions can be drawn for Poland from the USA experience of determining the range of health benefit basket? Irrespective of the similarities in the origin of health benefit basket development, both countries approached this issue differently. In the USA, the approach based on social dialogue and patient's perspective was selected while in Poland the perspective of public payer predominated. The transparency of principles and social dialogue constitute the fundamental elements of effective process of health benefit basket development and implementation which is both required and generally unpopular modification.

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

International Nuclear Information System (INIS)

D.R. Jackson; G.R. Kiebel

1999-01-01

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

Energy Technology Data Exchange (ETDEWEB)

D.R. Jackson; G.R. Kiebel

1999-08-24

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

Pandora's picnic basket: the potential and hazards of genetically modified foods

National Research Council Canada - National Science Library

McHughen, Alan

2000-01-01

... technologies in the them from with breeding whether examines and Picnic Basket subject, from potentially dangerous products. either use or about the regulatory processes established to protect us genetically other regulators modified "natural" the production. explains, in dear language, the technologies around the food, methods question of world of Researc...

High-resolution observations of combustion in heterogeneous surface fuels

Science.gov (United States)

E. Louise Loudermilk; Gary L. Achtemeier; Joseph J. O' Brien; J. Kevin Hiers; Benjamin S. Hornsby

2014-01-01

In ecosystems with frequent surface fires, fire and fuel heterogeneity at relevant scales have been largely ignored. This could be because complete burns give an impression of homogeneity, or due to the difficulty in capturing fine-scale variation in fuel characteristics and fire behaviour. Fire movement between patches of fuel can have implications for modelling fire...

Nucleoporins as components of the nuclear pore complex core structure and Tpr as the architectural element of the nuclear basket.

Science.gov (United States)

Krull, Sandra; Thyberg, Johan; Björkroth, Birgitta; Rackwitz, Hans-Richard; Cordes, Volker C

2004-09-01

The vertebrate nuclear pore complex (NPC) is a macromolecular assembly of protein subcomplexes forming a structure of eightfold radial symmetry. The NPC core consists of globular subunits sandwiched between two coaxial ring-like structures of which the ring facing the nuclear interior is capped by a fibrous structure called the nuclear basket. By postembedding immunoelectron microscopy, we have mapped the positions of several human NPC proteins relative to the NPC core and its associated basket, including Nup93, Nup96, Nup98, Nup107, Nup153, Nup205, and the coiled coil-dominated 267-kDa protein Tpr. To further assess their contributions to NPC and basket architecture, the genes encoding Nup93, Nup96, Nup107, and Nup205 were posttranscriptionally silenced by RNA interference (RNAi) in HeLa cells, complementing recent RNAi experiments on Nup153 and Tpr. We show that Nup96 and Nup107 are core elements of the NPC proper that are essential for NPC assembly and docking of Nup153 and Tpr to the NPC. Nup93 and Nup205 are other NPC core elements that are important for long-term maintenance of NPCs but initially dispensable for the anchoring of Nup153 and Tpr. Immunogold-labeling for Nup98 also results in preferential labeling of NPC core regions, whereas Nup153 is shown to bind via its amino-terminal domain to the nuclear coaxial ring linking the NPC core structures and Tpr. The position of Tpr in turn is shown to coincide with that of the nuclear basket, with different Tpr protein domains corresponding to distinct basket segments. We propose a model in which Tpr constitutes the central architectural element that forms the scaffold of the nuclear basket.

Preparation, Characterization and Adsorption Evaluation of old Newspaper Fibres using Basket Reactor (Nickel Removal by Adsorption)

International Nuclear Information System (INIS)

Ossman, M. E.; Abdelfatah, M.; Kiros, Y.

2016-01-01

In this work, old newspaper fibers bleached with H 2 O 2 , treated with KOH and treated with NaOCl were investigated as potential adsorbents. The characterization of the produced fibers using FTIR, SEM and particle size distribution have been carried out and tested for the removal of Ni (II) from aqueous solutions. The results indicated that the fibers treated with KOH give the highest %removal of Ni (II) with 88%. Two different reactor designs (batch and basket reactor) with different variables were studied. The results indicated that the equilibrium time was 30 min. and the removal of Ni (II) increased significantly as the p H increased from 2.0 to 6.0 and decreased at p H range of 6.5–8.0. The adsorption of Ni (II) onto old newspaper fibers treated with KOH using batch and basket reactors follows the Langmuir isotherm. The pseudo second order kinetic model provided good correlation for the adsorption of Ni (II) onto old newspaper fibers treated with KOH for both batch and basket reactors.

Evaluation of soil thermal potential under Tunisian climate using a new conic basket geothermal heat exchanger: Energy and exergy analysis

International Nuclear Information System (INIS)

Boughanmi, Hassen; Lazaar, Mariem; Farhat, Abdelhamid; Guizani, Amenallah

2017-01-01

Highlights: • Conic geothermal basket heat exchanger (CBGHE) is experimentally investigated. • Charging and discharging processes of CBGHE are evaluated. • Energy and exergy efficiencies of CBGHE are performed. • High and stable performance of surface geothermal energy in Tunisia is established. - Abstract: Geothermal heat exchangers system composed of two conic baskets serially connected is designed and realized. Both heat exchangers are made in polyethylene high-density material and have a length of 3 m each one. They will be used for greenhouse cooling and heating through a geothermal heat pump. Its conical geometry is selected to reduce the operation cost and the exploited area, compared to vertical and horizontal geothermal heat exchangers often used. It also assures the maximum of heat exchange with the soil. The aim of this study is to determine the thermal performance of one Conic Basket Geothermal Heat Exchanger (CBGHE), buried at 3 m deep, in the exploitation of the soil thermal potential, in summer. A rate of heat exchange with the soil is determined and the global heat exchange of the CBGHE is assessed. Its energy and exergy efficiencies are also evaluated using both first and second law of thermodynamic. Results show that the specific heat exchange ranges between 20 W m"−"1 and 50 W m"−"1. Maximal energetic and exergetic efficiencies of the CBGHE, equal to 62% and 37% respectively, are reached for a mass flow rate of 0.1 kg s"−"1. For this value of mass flow rate, the overall heat exchange coefficient is of 52 W m"−"2 K"−"1.

Plant basket hydraulic structures (PBHS) as a new river restoration measure.

Science.gov (United States)

Kałuża, Tomasz; Radecki-Pawlik, Artur; Szoszkiewicz, Krzysztof; Plesiński, Karol; Radecki-Pawlik, Bartosz; Laks, Ireneusz

2018-06-15

River restoration has become increasingly attractive worldwide as it provides considerable benefits to the environment as well as to the economy. This study focuses on changes of hydromorphological conditions in a small lowland river recorded during an experiment carried out in the Flinta River, central Poland. The proposed solution was a pilot project of the construction of vegetative sediment traps (plant basket hydraulic structures - PBHS). A set of three PBSH was installed in the riverbed in one row and a range of hydraulic parameters were recorded over a period of three years (six measurement sessions). Changes of sediment grain size were analysed, and the amount and size of plant debris in the plant barriers were recorded. Plant debris accumulation influencing flow hydrodynamics was detected as a result of the installation of vegetative sediment traps. Moreover, various hydromorphological processes in the river were initiated. Additional simulations based on the detected processes showed that the proposed plant basket hydraulic structures can improve the hydromorphological status of the river. Copyright © 2018 Elsevier B.V. All rights reserved.

Monitoring the Affordability of Healthy Eating: A Case Study of 10 Years of the Illawarra Healthy Food Basket

OpenAIRE

Williams, Peter

2010-01-01

Healthy food baskets have been used around the world for a variety of purposes, including: examining the difference in cost between healthy and unhealthy food; mapping the availability of healthy foods in different locations; calculating the minimum cost of an adequate diet for social policy planning; developing educational material on low cost eating and examining trends on food costs over time. In Australia, the Illawarra Healthy Food Basket was developed in 2000 to monitor trends in the af...

Monitoring the Affordability of Healthy Eating: A Case Study of 10 Years of the Illawarra Healthy Food Basket

Directory of Open Access Journals (Sweden)

Peter Williams

2010-11-01

Full Text Available Healthy food baskets have been used around the world for a variety of purposes, including: examining the difference in cost between healthy and unhealthy food; mapping the availability of healthy foods in different locations; calculating the minimum cost of an adequate diet for social policy planning; developing educational material on low cost eating and examining trends on food costs over time. In Australia, the Illawarra Healthy Food Basket was developed in 2000 to monitor trends in the affordability of healthy food compared to average weekly wages and social welfare benefits for the unemployed. It consists of 57 items selected to meet the nutritional requirements of a reference family of five. Bi-annual costing from 2000–2009 has shown that the basket costs have increased by 38.4% in the 10-year period, but that affordability has remained relatively constant at around 30% of average household incomes.

Monitoring the affordability of healthy eating: a case study of 10 years of the Illawarra Healthy Food Basket.

Science.gov (United States)

Williams, Peter

2010-11-01

Healthy food baskets have been used around the world for a variety of purposes, including: examining the difference in cost between healthy and unhealthy food; mapping the availability of healthy foods in different locations; calculating the minimum cost of an adequate diet for social policy planning; developing educational material on low cost eating and examining trends on food costs over time. In Australia, the Illawarra Healthy Food Basket was developed in 2000 to monitor trends in the affordability of healthy food compared to average weekly wages and social welfare benefits for the unemployed. It consists of 57 items selected to meet the nutritional requirements of a reference family of five. Bi-annual costing from 2000-2009 has shown that the basket costs have increased by 38.4% in the 10-year period, but that affordability has remained relatively constant at around 30% of average household incomes.

The corbiculate bees arose from New World oil-collecting bees: implications for the origin of pollen baskets.

Science.gov (United States)

Martins, Aline C; Melo, Gabriel A R; Renner, Susanne S

2014-11-01

The economically most important group of bees is the "corbiculates", or pollen basket bees, some 890 species of honeybees (Apis), bumblebees (Bombus), stingless bees (Meliponini), and orchid bees (Euglossini). Molecular studies have indicated that the corbiculates are closest to the New World genera Centris, with 230 species, and Epicharis, with 35, albeit without resolving the precise relationships. Instead of concave baskets, these bees have hairy hind legs on which they transport pollen mixed with floral oil, collected with setae on the anterior and middle legs. We sampled two-thirds of all Epicharis, a third of all Centris, and representatives of the four lineages of corbiculates for four nuclear gene regions, obtaining a well-supported phylogeny that has the corbiculate bees nested inside the Centris/Epicharis clade. Fossil-calibrated molecular clocks, combined with a biogeographic reconstruction incorporating insights from the fossil record, indicate that the corbiculate clade arose in the New World and diverged from Centris 84 (72-95)mya. The ancestral state preceding corbiculae thus was a hairy hind leg, perhaps adapted for oil transport as in Epicharis and Centris bees. Its replacement by glabrous, concave baskets represents a key innovation, allowing efficient transport of plant resins and large pollen/nectar loads and freeing the corbiculate clade from dependence on oil-offering flowers. The transformation could have involved a novel function of Ubx, the gene known to change hairy into smooth pollen baskets in Apis and Bombus. Copyright © 2014 Elsevier Inc. All rights reserved.

Surface and canopy fuels vary widely in 24-yr old postfire lodgepole pine forests

Science.gov (United States)

Nelson, K. N.; Turner, M.; Romme, W. H.; Tinker, D. B.

2013-12-01

Extreme fire seasons have become common in western North America, and the extent of young postfire forests has grown as fire frequency and annual area burned have increased. These young forests will set the stage for future fires, but an assessment of fuel loads in young forests is lacking. The rate of fuel re-accumulation and fuels variability in postfire forest landscapes is needed to anticipate future fire occurrence and behavior in the American West. We studied fuel characteristics in young lodgepole pine forests that regenerated after the 1988 fires in Yellowstone National Park to address two questions: (1) How do surface fuel characteristics change with time-since-fire? (2) How do canopy and surface fuels vary across the Yellowstone landscape 24 years postfire? During summer 2012, we re-measured surface fuels in 11 plots that were established in 1996 (8 yrs post fire), and we measured surface and canopy fuels in 82 stands (each 0.25 ha) distributed across the Yellowstone post-1988 fire landscape. In the remeasured plots, surface fuel loads generally increased over the last 16 years. One-hr fuels did not change between sample dates, but all other fuel classes (i.e., 10-hr, 100-hr, and 1000-hr) increased by a factor of two or three. Within the sample timeframe, variability of fuel loads within stands decreased significantly. The coefficients of variation decreased for all fuel classes by 23% to 67%. Data from the 82 plots revealed that canopy and surface fuels in 24-year-old stands varied tremendously across the Yellowstone landscape. Live tree densities spanned 0 to 344,067 trees ha-1, producing a mean available canopy fuel load of 7.7 Mg ha-1 and a wide range from 0 to 47 Mg ha-1. Total surface fuel loads averaged 130 Mg ha-1 and ranged from 49 to 229 Mg ha-1, of which 90% was in the 1000-hr fuel class. The mass of fine surface fuels (i.e., litter/duff, 1-hr, 10-hr, and herbaceous fuels) and canopy fuels (i.e., foliage and 1-hr branches) were strongly and

Novel method to achieve price-optimized, fully nutritious, health-promoting and acceptable national food baskets

DEFF Research Database (Denmark)

Parlesak, Alexandr; Robertson, Aileen

2015-01-01

available foods. The study was designed to obtain healthy, affordable, and socially acceptable diets for three European countries (Denmark, Slovenia, and Romania) and in three regions within Canada, Argentina, and Switzerland. Moreover, the costs for the “limiting” micronutrients and relative price......Purpose: The purpose of this study was to generate a framework for the development of health-promoting, fully nutritious, socially acceptable, and affordable national food baskets to be used as an advocacy tool by governments. In addition to containing all (micro-)nutrient requirements, food...... baskets should also reflect dietary guidelines to prevent non-communicable diseases and be optimized to achieve the highest possible social acceptance. So far, integrative approaches that include all these aspects are lacking. Methods: Food composition, local availability, food prices, national...

Basket call option pricing for CCVG using sparse grids

KAUST Repository

Crocce, Fabian

2016-01-06

The use of processes with jumps to overcome the shortcomings of the classical Black and Scholes when modelling stock prices has became very popular. One of the best-known models is the Common Clock Variance Gamma model (CCVG), introduced by Madan and Seneta in the 1990 [3]. We propose a method to price European basket call options modelled by the CCVG. The method could be extended to other model obtained by the subordination of a multidimensional Brownian motion and to more general options. To simplify the expositions we consider calls under the CCVG.

Basket call option pricing for CCVG using sparse grids

KAUST Repository

Crocce, Fabian; Hä ppö lä , Juho; Iania, Alessandro; Tempone, Raul

2016-01-01

The use of processes with jumps to overcome the shortcomings of the classical Black and Scholes when modelling stock prices has became very popular. One of the best-known models is the Common Clock Variance Gamma model (CCVG), introduced by Madan and Seneta in the 1990 [3]. We propose a method to price European basket call options modelled by the CCVG. The method could be extended to other model obtained by the subordination of a multidimensional Brownian motion and to more general options. To simplify the expositions we consider calls under the CCVG.

Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

International Nuclear Information System (INIS)

1990-02-01

The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs

A Nutritional Analysis of the Food Basket in BIH: A Linear Programming Approach

Directory of Open Access Journals (Sweden)

Arnaut-Berilo Almira

2017-04-01

Full Text Available This paper presents linear and goal programming optimization models for determining and analyzing the food basket in Bosnia and Herzegovina (BiH in terms of adequate nutritional needs according to World Health Organization (WHO standards and World Bank (WB recommendations. A linear programming (LP model and goal linear programming model (GLP are adequate since price and nutrient contents are linearly related to food weight. The LP model provides information about the minimal value and the structure of the food basket for an average person in BiH based on nutrient needs. GLP models are designed to give us information on minimal deviations from nutrient needs if the budget is fixed. Based on these results, poverty analysis can be performed. The data used for the models consisted of 158 food items from the general consumption of the population of BiH according to COICOP classifications, with average prices in 2015 for these products.

River channel morphology and hydraulics properties due to introduction of plant basket hydraulic structures for river channel management

Science.gov (United States)

Kałuża, Tomasz; Radecki-Pawlik, Artur; Plesiński, Karol; Walczak, Natalia; Szoszkiewicz, Krzysztof; Radecki-Pawlik, Bartosz

2016-04-01

In the present time integrated water management is directly connected with management and direct works in river channels themselves which are taking into account morphological processes in rivers and improve flow conditions. Our work focused on the hydraulic and hydrodynamic consequences upon the introduction of the concept of the improvement of the hydromorphological conditions of the Flinta River in a given reach following river channel management concept. Based on a comprehensive study of the hydromorphological state of the river, four sections were selected where restoration measures can efficiently improve river habitat conditions in the river. For each section a set of technical and biological measures were proposed and implemented in practice. One of the proposed solutions was to construct plant basket hydraulic structures (PBHS) within the river channel, which are essentially plant barriers working as sediment traps, changing river channel morphology and are in line with concepts of Water Framework Directive. These relatively small structures work as crested weirs and unquestionably change the channel morphology. Along our work we show the results of three-year long (2013-2015) systematic measurements that provided information on the morphological consequences of introducing such structures into a river channel. Our main conclusions are as follows: 1. Plant basket hydraulic structures cause changes in hydrodynamic conditions and result in sediment accumulation and the formation of river backwaters upstream and downstream the obstacle; 2. The introduced plant basket hydraulic structures cause plant debris accumulation which influences the hydrodynamic flow conditions; 3. The installation of plant basket hydraulic structures on the river bed changes flow pattern as well as flow hydrodynamic conditions causing river braiding process; 4. The erosion rate below the plant basket hydraulic structures is due to the hydraulic work conditions of the PBHS and its

Hanford spent nuclear fuel project update

Energy Technology Data Exchange (ETDEWEB)

Williams, N.H.

1997-08-19

Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building.

LOFT fuel rod surface temperature measurement testing

International Nuclear Information System (INIS)

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

1978-01-01

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

Alternative concepts for spent fuel storage basin expansion at Morris Operation

International Nuclear Information System (INIS)

Graf, W.A. Jr.; King, C.E.; Miller, G.P.; Shadel, F.H.; Sloat, R.J.

1980-08-01

Alternative concepts for increasing basin capabilities for storage of spent fuel at the Morris Operation have been defined in a series of simplified flow diagrams and equipment schematics. Preliminary concepts have been outlined for (1) construction alternatives for an add-on basin, (2) high-density baskets for storage of fuel bundles or possible consolidated fuel rods in the existing or add-on basins, (3) modifications to the existing facility for increasing cask handling and fuel receiving capabilities and (4) accumulation, treatment and disposal of radwastes from storage operations. Preliminary capital and operating costs have been prepared and resource and schedule requirements for implementing the concepts have been estimated. The basin expansion alternatives would readily complement potential dry storage projects at the site in an integrated multi-stage program that could provide a total storage capacity of up to 7000 tonnes of spent fuel

Percutaneous Retrieval of a Dislodged Guide Wire by Dormia Basket in a 98-year-old Patient, Still a Safe and Feasible Approach

Directory of Open Access Journals (Sweden)

Gh Soltani

2010-06-01

Full Text Available Since the first report of percutaneous retrieval of intravascular foreign body in 1964, it has been accepted as a favorite approach for intravascular foreign body removal. Various instruments such as snares, biopsy forceps, dormia basket or tip deflecting wires are available for this approach. Herein, we report percutaneous retrieval of a dislodged guide-wire by dormia basket. The Patient was a 98-year-old female who was admitted in the Intensive care unit due to confusional state. The physian incharge left the guide-wire in her subclavian vein during central venous catheter insertion. He realized his fault immediately, so he started anti-coagulation therapy and requested surgical consult. Because of the patient’s high risk of surgery, the surgeon referred the patient to cardiologist for non-invasive guide wire removal. The guide wire was apprpached through femoral vein by a dormia basket (4/6 F and it was removed without any difficulties or complications. Utilization of baskets for intravascular foreign body removal is frequently reported and has been successful with low complication rates; In addition, the low cost of the device makes it less of a burden to the patient and the hospital.

Use of Linear Programming to Develop Cost-Minimized Nutritionally Adequate Health Promoting Food Baskets

DEFF Research Database (Denmark)

Parlesak, A.; Tetens, Inge; Dejgård Jensen, Jørgen

2016-01-01

Food-Based Dietary Guidelines (FBDGs) are developed to promote healthier eating patterns, but increasing food prices may make healthy eating less affordable. The aim of this study was to design a range of cost-minimized nutritionally adequate health-promoting food baskets (FBs) that help prevent ...

Validation of the high performance leadership competencies as measured by an assessment centre in-basket

Directory of Open Access Journals (Sweden)

H. H. Spangenberg

2003-10-01

Full Text Available The purpose of this study was to validate Schroder’s High Performance Leadership Competencies (HPLCs, measured by a specially designed In-basket, against multiple criteria. These consisted of six measures of managerial success, representing managerial advancement and salary progress criteria, and a newly developed comprehensive measure of work unit performance, the Performance Index. An environmental dynamism and complexity questionnaire served as moderator variable. Results indicated disappointing predictive validity quotients for the HPLCs as measured by an In-basket, in contrast to satisfactory predictive and construct validity obtained in previous studies by means of a full assessment centre. The implications of the findings are discussed and suggestions are made for improving the validity of the In-basket. Opsomming Die doel van hierdie studie was die validering van Schroder se Hoëvlak Leierskapsbevoegdhede, gemeet deur ‘n spesiaal ontwerpte Posmandjie, teen veelvoudige kriteria. Dit behels ses metings van bestuursukses wat bestuursbevorderings- en salarisvorderingskriteria insluit, sowel as ‘n nuutontwikkelde, omvattende meting van werkeenheidsprestasie, die Prestasie indeks. ‘n Vraelys wat die dinamika en kompleksiteit van die omgewing meet, het as moderator veranderlike gedien. Resultate dui op teleurstellende geldigheidskwosiënte vir die Hoëvlak Leierskapsbevoegdhede soos gemeet deur ‘n posmandjie, in teenstelling met bevredigende voorspellings- en konstrukgeldigheid wat in vorige studies deur middel van ‘n volle takseersentrum verkry is. Die bevindinge word bespreek en voorstelle word gemaak om die geldigheidskwosiënte te verbeter.

Design characteristics of pantograph type in vessel fuel handling system in SFR

International Nuclear Information System (INIS)

Kim, S. H.; Koo, G. H.

2012-01-01

The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

Design characteristics of pantograph type in vessel fuel handling system in SFR

Energy Technology Data Exchange (ETDEWEB)

Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

2012-10-15

The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

Regulation of the Hippocampal Network by VGLUT3-Positive CCK- GABAergic Basket Cells

Directory of Open Access Journals (Sweden)

Caroline Fasano

2017-05-01

Full Text Available Hippocampal interneurons release the inhibitory transmitter GABA to regulate excitation, rhythm generation and synaptic plasticity. A subpopulation of GABAergic basket cells co-expresses the GABA/glycine vesicular transporters (VIAAT and the atypical type III vesicular glutamate transporter (VGLUT3; therefore, these cells have the ability to signal with both GABA and glutamate. GABAergic transmission by basket cells has been extensively characterized but nothing is known about the functional implications of VGLUT3-dependent glutamate released by these cells. Here, using VGLUT3-null mice we observed that the loss of VGLUT3 results in a metaplastic shift in synaptic plasticity at Shaeffer’s collaterals – CA1 synapses and an altered theta oscillation. These changes were paralleled by the loss of a VGLUT3-dependent inhibition of GABAergic current in CA1 pyramidal layer. Therefore presynaptic type III metabotropic could be activated by glutamate released from VGLUT3-positive interneurons. This putative presynaptic heterologous feedback mechanism inhibits local GABAergic tone and regulates the hippocampal neuronal network.

A new framework to assess risk for a spent fuel dry storage facility

International Nuclear Information System (INIS)

Ryu, J. H.; Jae, M. S.; Jung, C. W.

2004-01-01

A spent fuel dry storage facility is a dry cooling storage facility for storing irradiated nuclear fuel and associated radioactive materials. It has very small possibilities to release radiation materials. It means a safety analysis for a spent fuel dry storage facility is required before construction. In this study, a new framework for assessing risk associated with a spent fuel dry storage facility is represented. A safety assessment framework includes 3 modules such as assessment of basket/cylinder failure rates, that of overall storage system, and site modeling. A reliability physics model for failure rates, event tree analysis(ETA)/fault tree analysis for system analysis, Bayesian analysis for initial events data, and MACCS code for consequence analysis have been used in this study

Cummings Memorial Lecture - 1975. The market basket: food for thought.

Science.gov (United States)

Deichman, W B

1975-06-01

The world food crisis is as critical today as when it was debated at the 1974 World Food Conference in Rome. Since the United States and Canada-and to a lesser extent, Australia and New Zealand-lead in the production of corn, wheat and soybeans, the North American "bread basket" has become the "market basket" of the world. For welfare, economic, and political reasons, our energies, resources, and deliberations must be expanded toward optimum production of wholesome food products. I do not recommend that we permit food additives in "questionably" safe or excessive concentrations in our agricultural products. I do recommend, however, that tolerance limits for food additives be established based on a comprehensive review of all contributing factors-the world food crisis and the rational interpretation of both positive and negative animal data as they relate to man. As Dr. Herbert Stokinger put it so aptly: "Avoid the establishment of unnecessarily severe standards." 2. Funds for research and teaching of food and nutrition should be greatly increased, so that all who can read and write may be made aware of the daily dietary requirements for the maintenance of good health. 3. Unsubstantiated scare tactics in publications of the scientific and lay press can only lead to well-intended but often emotionally-inspired restrictions, ordinances, and laws. Such decisions are likely to either under- or over-define the requirements and standards for food additives and other chemicals which are important to the well-being of the populace.

Thermal analysis of dry concrete canister storage system for CANDU spent fuel

International Nuclear Information System (INIS)

Ryu, Yong Ho

1992-02-01

This paper presents the results of a thermal analysis of the concrete canisters for interim dry storage of spent, irradiated Canadian Deuterium Uranium(CANDU) fuel. The canisters are designed to contain 6-year-old fuel safely for periods of 50 years in stainless steel baskets sealed inside a steel-lined concrete shield. In order to assure fuel integrity during the storage, fuel rod temperature shall not exceed the temperature limit. The contents of thermal analysis include the following : 1) Steady state temperature distributions under the conservative ambient temperature and insolation load. 2) Transient temperature distributions under the changes in ambient temperature and insolation load. Accounting for the coupled heat transfer modes of conduction, convection, and radiation, the computer code HEATING5 was used to predict the thermal response of the canister storage system. As HEATING5 does not have the modeling capability to compute radiation heat transfer on a rod-to-rod basis, a separate calculating routine was developed and applied to predict temperature distribution in a fuel bundle. Thermal behavior of the canister is characterized by the large thermal mass of the concrete and radiative heat transfer within the basket. The calculated results for the worst case (steady state with maximum ambient temperature and design insolation load) indicated that the maximum temperature of the 6 year cooled fuel reached to 182.4 .deg. C, slightly above the temperature limit of 180 .deg. C. However,the thermal inertia of the thick concrete wall moderates the internal changes and prevents a rise in fuel temperature in response to ambient changes. The maximum extent of the transient zone was less than 75% of the concrete wall thickness for cyclic insolation changes. When transient nature of ambient temperature and insolation load are considered, the fuel temperature will be a function of the long term ambient temperature as opposed to daily extremes. The worst design

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

International Nuclear Information System (INIS)

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

1997-09-01

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

Many-core architectures boost the pricing of basket options on adaptive sparse grids

DEFF Research Database (Denmark)

Heinecke, Alexander; Jepsen, Jacob; Bungartz, Hans Joachim

2013-01-01

of the operator's parallelism, we are able to achieve nearly perfect strong scaling for the Black-Scholes solver. Our results show that typical problem sizes (5 dimensional basket options), require at least 4 NVIDIA K20X Kepler GPUs (inside a Cray XK7) in order to be faster than the Up-Down scheme running on 16...

Multiple and substitute addictions involving prescription drugs misuse among 12th graders: gateway theory revisited with Market Basket Analysis.

Science.gov (United States)

Jayawardene, Wasantha Parakrama; YoussefAgha, Ahmed Hassan

2014-01-01

This study aimed to identify the sequential patterns of drug use initiation, which included prescription drugs misuse (PDM), among 12th-grade students in Indiana. The study also tested the suitability of the data mining method Market Basket Analysis (MBA) to detect common drug use initiation sequences in large-scale surveys. Data from 2007 to 2009 Annual Surveys of Alcohol, Tobacco, and Other Drug Use by Indiana Children and Adolescents were used for this study. A close-ended, self-administered questionnaire was used to ask adolescents about the use of 21 substance categories and the age of first use. "Support%" and "confidence%" statistics of Market Basket Analysis detected multiple and substitute addictions, respectively. The lifetime prevalence of using any addictive substance was 73.3%, and it has been decreasing during past few years. Although the lifetime prevalence of PDM was 19.2%, it has been increasing. Males and whites were more likely to use drugs and engage in multiple addictions. Market Basket Analysis identified common drug use initiation sequences that involved 11 drugs. High levels of support existed for associations among alcohol, cigarettes, and marijuana, whereas associations that included prescription drugs had medium levels of support. Market Basket Analysis is useful for the detection of common substance use initiation sequences in large-scale surveys. Before initiation of prescription drugs, physicians should consider the adolescents' risk of addiction. Prevention programs should address multiple addictions, substitute addictions, common sequences in drug use initiation, sex and racial differences in PDM, and normative beliefs of parents and adolescents in relation to PDM.

Real purchasing power of oil revenues for OPEC Member Countries: a broad currency basket and dynamic trade pattern approach

International Nuclear Information System (INIS)

Mazraati, M.

2005-01-01

The purpose of this paper is to examine the real purchasing power of OPEC Member Countries' oil revenues, which are subject to 'the value of the US dollar vis-a-vis other major currencies' and 'world imported inflation'. The exponential weighting average formula with a broad basket of currencies is suggested. The basket of currencies is labelled as a broad currency basket and includes the major trading partners of OPEC Member Countries. The weights are normalized OPEC import shares of the countries of the basket and are updated and adjusted every year to incorporate a gradual change in the trade pattern. In other words, the dynamic trade pattern approach is incorporated in the calculations. The nominal dollar oil revenues of OPEC Member Countries are about $5,099 billion during 1970 to 2004, of which $3,725 bn (73 per cent) have been lost due to imported inflation and the dollar's depreciation. Imported inflation and dollar depreciation have had a respective 78.6 per cent and 21.4 per cent contribution to the losses of the purchasing power of OPEC Member Countries. The imported inflation rate approaches a stable low level, but OPEC still has a lot of concerns on dollar swings. The euro offers opportunities for many oil-exporting nations that have extensive trade relations with Euro-zone countries. Payments for oil exports can be invoked in euros at the prevailing dollar-euro rate on the day of a given contract, or any other trigger formula. This would immunize a major portion of OPEC oil revenues from dollar depreciation. (author)

Development and testing of metallic fuels with high minor actinide content

International Nuclear Information System (INIS)

Meyer, M.K.; Hayes, S.L.; Kennedy, J.R.; Keiser, D.D.; Hilton, B.A.; Frank, S.M.; Kim, Y.-S.; Chang, G.; Ambrosek, R.G.

2003-01-01

Metallic alloys are promising candidates for use as fuels for transmutation and in advanced closed nuclear cycles. Metallic alloys have high heavy metal atom density, relatively high thermal conductivity, favorable gas release behavior, and lend themselves to remote recycle processes. Both non-fertile and uranium-bearing metal fuels containing minor actinide are under consideration for use as transmutation fuels by the U.S. Advanced Fuel Cycle (AFC) program, however, little irradiation performance data exists for fuel forms containing significant fractions of minor actinides. The first irradiation tests of non-fertile high-actinide-content fuels are scheduled to begin in early 2003 in the Advanced Test Reactor (ATR). The irradiation test matrix was designed to provide basic information on the irradiation behavior of binary Pu-Zr alloy fuel and the effect of the minor actinides americium and neptunium on alloy fuel behavior, together and separately. Five variants of transuranic containing zirconium-based alloy fuels are included in the AFC-1 irradiation test matrix. These are (in wt.%) Pu-40Zr, Pu-60Zr, Pu-12Am-40Zr, Pu-10Np-40Zr and Pu-10Np-10Am-40Zr. PuN-ZrN based fuels containing Am and Np are also included. All five of the fuel alloys have been fabricated in the form of cylindrical fuel slugs by arc-casting. Short melt times, on the order or 5-20 seconds, prevent the volatilization of significant quantities of americium metal, despite the high melt temperatures characteristic of the arc-melting process. Alloy microstructure have been characterized by x-ray diffraction and scanning electron microscopy. Thermal analysis has also been performed. The AFC-1 irradiation experiment configuration consists of twenty-four sodium bonded fuel specimens sealed in helium filled secondary capsules. The first capsule has a design burnup to 7 at.% 239 Pu; goal peak burnup of the second capsule is ∼18 at%. Capsule assemblies are placed within an aluminum flow-through basket

Training simulator for teaching a technique to the long transmission of ball in basket-ball by a method by an arcuated hand from above with threaten

Directory of Open Access Journals (Sweden)

Charikova K.M.

2012-03-01

Full Text Available Technical devices which used in basket-ball are considered. The features of constructing of trainers and method of their application are selected in a training process. A trainer is offered for teaching a technique to the long transmission of ball in basket-ball. A trainer is a moving on a rope imitator of basket-ball ball. This construction allows to design initial position, замах for implementation of transmission and line of acceleration of ball in the final phase of motion. The method of the use of trainer is developed in an educational process.

Kinetic enantioselectivity of a protonated bis(diamido)-bridged basket resorcin[4]arene towards alanine peptides.

Science.gov (United States)

Fraschetti, C; Montagna, M; Crestoni, M E; Calcaterra, A; Aiello, F; Santi, L; Filippi, A

2017-02-01

Efficient enantiodiscrimination of some alanine-containing di- and tri-peptides by using chiral protonated bis(diamido)-bridged basket resorcin[4]arenes depends on several factors, including the basicity of the amino acid residues at the C- and N-termini of the peptide.

Characterization of Hanford K basin spent nuclear fuel and sludge

International Nuclear Information System (INIS)

Lawrence, L.A.

1996-01-01

A characterization plan was prepared to support the Integrated Process Strategy (IPS) for resolution of the safety and environmental concerns associated with the deteriorating Spent Nuclear Fuel (SNF) stored in the Hanford Site K Basins. This plan provides the structure and logic and identifies the information needs to be supported by the characterization activities. The IPS involves removal of the fuel elements from the storage canister and placing them in a container, i.e., Multiple Canister Overpack (MCO) capable of holding multiple tiers of baskets full of fuel. The MCOs will be vacuum dried to remove free water and shipped to the Container Storage Building (CSB) where they will be staged waiting for hot vacuum conditioning. The MCO will be placed in interim storage in the CSB following conditioning and disposition

APLIKASI DATA MINING MARKET BASKET ANALYSIS PADA TABEL DATA ABSENSI ELEKTRONIK UNTUK MENDETEKSI KECURANGAN ABSENSI (CHECK-LOCK KARYAWAN DI PERUSAHAAN

Directory of Open Access Journals (Sweden)

Gregorius Satia Budhi

2007-01-01

Full Text Available Taking attendance from employees always becomes a problem for Human Resource Department (HRD in many companies lately. Although there is an automatic check-lock machine, it still has a weakness. This machine can't detect some frauds like the employee swipes double identity card, his card and the others card. Reseachers want to solve this problem by using data mining method, especially market basket analysis.This software will transform the attendance data to compact transaction format by using MaxDiff Histogram method. And it will be processed into frequent itemset with Pincer Search Algoritm. At the final process the employee's association rule will got from frequent itemset. This output will be served to user that is the HRD of a firm.Testing result shows that Data Mining Market Basket Analysis can be used to get pattern of employee's check-lock from a company. And this pattern can help user to detect fraud that is done by employee. Abstract in Bahasa Indonesia : Absensi pegawai selama ini selalu menjadi permasalahan yang pelik bagi bagian HRD di perusahaan - perusahaan yang ada. Walaupun telah ada peralatan absensi otomatis, alat ini masih memiliki kelemahan yaitu, tidak dapat mendeteksi kecurangan pegawai untuk menitipkan kartu absensinya pada karyawan lain untuk diabsenkan. Peneliti berkeinginan untuk mengatasi permasalahan absensi tersebut dengan memanfaatkan metode data mining, khususnya metode market basket analysis, untuk mendeteksi kecurangan ini.Perangkat lunak yang dibuat ini akan mentranformasikan data absensi pegawai menggunakan metode MaxDiff Histogram menjadi format compact transaction yang selanjutnya akan diproses menggunakan Algoritma Pincer Search menjadi frequent itemset. Pada akhirnya dari data frequent itemset ini didapat association rule pegawai untuk disajikan kepada pengguna, yaitu bagian HRD perusahaan.Dari hasil pengujian dapat diketahui bahwa metode Data Mining Market Basket Analysis dapat dimanfaatkan untuk menggali

SPE (tm) regenerative hydrogen/oxygen fuel cells for extraterrestrial surface and microgravity applications

Science.gov (United States)

Mcelroy, J. F.

1990-01-01

Viewgraphs on SPE regenerative hydrogen/oxygen fuel cells for extraterrestrial surface and microgravity applications are presented. Topics covered include: hydrogen-oxygen regenerative fuel cell energy storage system; electrochemical cell reactions; SPE cell voltage stability; passive water removal SPE fuel cell; fuel cell performance; SPE water electrolyzers; hydrophobic oxygen phase separator; hydrophilic/electrochemical hydrogen phase separator; and unitized regenerative fuel cell.

Application of fuel cells in surface ships

Energy Technology Data Exchange (ETDEWEB)

Bourne, C.; Nietsch, T.; Griffiths, D.; Morley, J.

2001-07-01

This report presents the findings of a DTI supported project entitled: ''Applications of fuel cells in surface ships''. It gives a brief market analysis describing the general requirements of different vessel types and an overview of the different heat engine technologies currently used for propulsion and power generation in ships. The appendices contain a more detailed description of the different vessel types, their general requirements and a description of current prime mover technologies used. This analysis is followed by a summary of the major fuel cell development programmes and activities ongoing in different countries that have a direct or potential relevance to a marine application of the technology. (author)

Equivalent thermal conductivity of the storage basket with spent nuclear fuel of VVER-1000 reactors

International Nuclear Information System (INIS)

Alyokhina, Svitlana; Kostikov, Andriy

2014-01-01

Due to limitation of computation resources and/or computation time many thermal problems require to use simplified geometrical models with equivalent thermal properties. A new method for definition of equivalent thermal conductivity of spent nuclear fuel storage casks is proposed. It is based on solving the inverse heat conduction problem. For the proposed method two approaches for equivalent thermal conductivity definition were considered. In the first approach a simplified model in conjugate formulation is used, in the second approach a simplified model of solid body which allows an analytical solution is used. For safety ensuring during all time of spent nuclear fuel storage the equivalent thermal conductivity was calculated for different storage years. The calculated equivalent thermal conductivities can be used in thermal researches for dry spent nuclear fuel storage safety.

Design of a stationary-basket (microreactor)

Energy Technology Data Exchange (ETDEWEB)

Timidei, A; Raisa, L

1977-05-01

A continuous-flow stirred-tank microreactor (295 cc capacity) for kinetic studies in gas-solid heterogeneous catalysis has been designed to obtain an apparatus characterized by perfect mixing, absence of resistance to mass and heat transfer, and pressure tightness, and which would not require sophisticated machine-shop work. This reactor is suitable for operation at up to 300/sup 0/C and 20 atm, and has clear advantages over the spinning-basket or rotating reactor in studies of catalysts of low mechanical strength. Tests with nitrogen or ethane as tracer and helium as carrier gas, run with a thermal conductivity detector, gave a ''decadence'' curve with a -0.97 slope at all mixing speeds, compared with a -1.0 slope for perfect mixing. Thermal gradients between fluid and pellet were minimal for three different pellet positions within the annular bed, at 92/sup 0/C. In benzene hydrogenation over a nickel/kieselguhr catalyst, measurement of reaction rates as a function of mixing speeds showed that good results in kinetic studies were obtainable at 1500 rpm and above. Diagrams and graphs.

Quantum mechanical design and structure of the Li@B10H14 basket with a remarkably enhanced electro-optical response.

Science.gov (United States)

Muhammad, Shabbir; Xu, Hongliang; Liao, Yi; Kan, Yuhe; Su, Zhongmin

2009-08-26

An innovative type of lithium decahydroborate (Li@B(10)H(14)) complex with a basketlike complexant of decaborane (B(10)H(14)) has been designed using quantum mechanical methods. As Li atom binds in a handle fashion to terminal electrophilic boron atoms of the decaborane basket, its NBO charge q (Li) is found to be 0.876, close to +1. This shows that the Li atom has been ionized to form a cation and an anion at the open end of B(10)H(14). The most fascinating feature of this Li doping is its loosely bound valence electron, which has been pulled into the cavity of the B(10)H(14) basket and become diffuse by the electron-deficient morphological features of the open end of the B(10)H(14) basket. Strikingly, the first hyperpolarizability (beta(0)) of Li@B(10)H(14) is about 340 times larger than that of B(10)H(14), computed to be 23,075 au (199 x 10(-30) esu) and 68 au, respectively. Besides this, the intercalation of the Li atom to the B(10)H(14) basket brings some distinctive changes in its Raman, (11)B NMR, and UV-vis spectra along with its other electronic properties that might be used by the experimentalists to identify this novel kind of Li@B(10)H(14) complex with a large electro-optical response. This study may evoke the possibility to explore a new thriving area, i.e., alkali metal-boranes for NLO application.

Hanford K basins spent nuclear fuel project update

International Nuclear Information System (INIS)

Williams, N.H.; Hudson, F.G.

1997-07-01

Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building

Selenium fuel: Surface engineering of U(Mo) particles to optimise fuel performance

International Nuclear Information System (INIS)

Van den Berghe, S.; Leenaers, A.; Detavernier, C.

2010-01-01

Recent developments on the stabilisation of U(Mo) in-pile behaviour in plate-type fuel have focussed almost exclusively on the addition of Si to the Al matrix of the fuel. This has now culminated in a qualification effort in the form of the European LEONIDAS initiative for which irradiations will start in 2010. In this framework, many discussions have been held on the Si content of the matrix needed for stabilisation of the interaction phase and the requirement for the formation of Si-rich layers around the particles during the fabrication steps. However, it is clear that the Si needs to be incorporated in the interaction phase for it to be effective, for which the currently proposed methods depend on a diffusion mechanism, which is difficult to control. This has lead to the concept of a Si coated particle as a more efficient way of incorporating the Si in the fuel by putting it immediately where it will be required : at the fuel-matrix interface. As part of the SELENIUM (Surface Engineered Low ENrIched Uranium-Molybdenum fuel) project, SCK CEN has built a sputter coater for PVD magnetron sputter coating of particles in collaboration with the University of Ghent. The coater is equipped with three 3 inch magnetron sputter heads, allowing deposition of 3 different elements or a single element at high deposition speed. The particles are slowly rotated in a drum to produce homogeneous layer thicknesses. (author)

Characterization of internal surface finishing of tubes for CAREM 25 fuel rods

International Nuclear Information System (INIS)

Loureiro, N.V; Juarez, G; Bianchi, D; Flores, A; Vizcaino, P

2012-01-01

One of the factors that ensure the good behavior of the fuel claddings of the nuclear power reactors is the internal surface quality. In the present work has been carried out a study of the internal surface of the tube after a cold rolling process developed in the Departamento de Tecnologia de Aleaciones de Circonio and applied by FAE-SA and PPFAE-CNEA in each rolling stage to obtain the fuel claddings for the reactor CAREM 25. The inner surface has been observed by scanning electron microscopy, SEM, being the objective of this study to verify not only the good internal surface but also infer about how starting from tubes of different initial diameter reduction the quality of the final product will be affected. The manufacturing process of the tubes for this new fuel went through modifications during the development, adding intermediate chemical pickling stages in order to improve the internal surface quality of the final product. From determinations made with ultrasound, the defects charts obtained made it possible to compare the observed signals more relevant and the micrographs in these areas in order to characterize possible defects (author)

Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology

Directory of Open Access Journals (Sweden)

Eun-Young Choi

2015-12-01

Full Text Available The electrochemical reduction process has been used to reduce spent oxide fuel to a metallic form using pyroprocessing technology for a closed fuel cycle in combination with a metal-fuel fast reactor. In the electrochemical reduction process, oxides fuels are loaded at the cathode basket in molten Li2O–LiCl salt and electrochemically reduced to the metal form. Various approaches based on thermodynamic calculations and experimental studies have been used to understand the electrode reaction and efficiently treat spent fuels. The factors that affect the speed of the electrochemical reduction have been determined to optimize the process and scale-up the electrolysis cell. In addition, demonstrations of the integrated series of processes (electrorefining and salt distillation with the electrochemical reduction have been conducted to realize the oxide fuel cycle. This overview provides insight into the current status of and issues related to the electrochemical processing of spent nuclear fuels.

Dynamic Effect of a Change in the Exchange Rate System: From a Fixed Regime to a Basket-Peg or a Floating Regime

OpenAIRE

Yoshino, Naoyuki; Kaji, Sahoko; Asonuma, Tamon

2015-01-01

This paper theoretically evaluates the dynamic effects of a shift in an exchange rate system from a fixed regime to a basket peg, or to a floating regime, and obtains transition paths for the shift based on a dynamic stochastic general equilibrium model of a small open economy. We apply quantitative analysis using data from the People's Republic of China and Thailand and find that a small open country would be better off shifting to a basket peg or to a floating regime than maintaining a doll...

Ru-decorated Pt surfaces as model fuel cell electrocatalysts for CO electrooxidation.

Science.gov (United States)

Maillard, F; Lu, G-Q; Wieckowski, A; Stimming, U

2005-09-01

This feature article concerns Pt surfaces modified (decorated) by ruthenium as model fuel cell electrocatalysts for electrooxidation processes. This work reveals the role of ruthenium promoters in enhancing electrocatalytic activity toward organic fuels for fuel cells, and it particularly concerns the methanol decomposition product, surface CO. A special focus is on surface mobility of the CO as it is catalytically oxidized to CO(2). Different methods used to prepare Ru-decorated Pt single crystal surfaces as well as Ru-decorated Pt nanoparticles are reviewed, and the methods of characterization and testing of their activity are discussed. The focus is on the origin of peak splitting involved in the voltammetric electrooxidation of CO on Ru-decorated Pt surfaces, and on the interpretative consequences of the splitting for single crystal and nanoparticle Pt/Ru bimetallic surfaces. Apparently, screening through the literature allows formulating several models of the CO stripping reaction, and the validity of these models is discussed. Major efforts are made in this article to compare the results reported by the Urbana-Champaign group and the Munich group, but also by other groups. As electrocatalysis is progressively more and more driven by theory, our review of the experimental findings may serve to summarize the state of the art and clarify the roads ahead. Future studies will deal with highly dispersed and reactive nanoscale surfaces and other more advanced catalytic materials for fuel cell catalysis and related energy applications. It is expected that the metal/metal and metal/substrate interactions will be increasingly investigated on atomic and electronic levels, with likewise increasing participation of theory, and the structure and reactivity of various monolayer catalytic systems involving more than two metals (that is ternary and quaternary systems) will be interrogated.

Research of level of formed knowledge, abilities and skills on basket-ball for the students of faculty of physical education.

Directory of Open Access Journals (Sweden)

Kovalenko J.A.

2011-06-01

Full Text Available The results of a zero control of students of faculty of physical education are presented. The estimation of theoretical knowledge's is conducted, abilities and skills on basket-ball for the students of a 1 course of directions of preparation «Physical education», «Health of man» and «Sport». It is set that the level of their formed at most students of different directions of preparation is insufficient. It is certain that by reason of such position, foremost, there is insufficient preparation and playing practice of students during teaching at general school. A diary is developed on basket-ball for student general schools.

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

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

International Nuclear Information System (INIS)

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

1988-10-01

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

Percutaneous treatment of thrombosed native arteriovenous dialysis fistula insufficiency: efficacy of mechanical thrombectomy with using the stone basket

International Nuclear Information System (INIS)

Kim, Young Hwan; Ko, Sung Min; Kim, Mi Jung; Kwon, Jung Hyeok; Sohn, Cheol Ho; Choi, Jin Soo; Park, Kyung Sik; Kim, Yong Joo

2006-01-01

We wanted to evaluate the procedural success after percutaneous treatment of thrombosed native arteriovenous dialysis fistula insufficiency and the efficacy of performing mechanical thrombectomy with using the stone basket. From March 2004 to June 2005, 36 thrombosed native hemodialysis access shunts in the upper limbs (brachiocephalic fistulas: 16 and radiocephalic fistulas: 20) were percutaneously treated in 30 patients. Declotting procedures were performed with using urokinase (100,00-200,000 unit) and manual catheter-directed thrombo-aspiration in all the patients. Angioplasty (6 mm in diameter and 4 cm in length) was performed at the identified area of the stenosis and /or with maceration of the thrombus. In 14 cases with massive thrombosis that was refractory to the above mentioned declotting procedures, mechanical thrombectomy with using a Wittich nitinol stone basket (Cook, Bloomington, IN) was performed. Data regarding the procedural success rate and the patency rate were analyzed by means of Fischer's exact test, and the Kaplan-Meier method with the Log-rank test was used for statistical inter-group comparisons between the brachiocephalic and radiocephalic fistulas. Successful declotting and restoration of thrill were achieved in 30 of 36 procedures (83%). Reestablishment of normal dialysis for at least one session was achieved in 29 of 36 procedures (81%). The procedural success rate for the brachiocephalic fistulas was 94% compared with 70% for the radiocephalic fistulas, but the difference was not statistically significant (ρ = 0.104). In the cases with performing mechanical thrombectomy and using the stone basket, procedural success was achieved in 93% (13/14). The expected patency rates at 3, 6 and 12 months were 78%, 61% and 51%, respectively. The patency rates after declotting procedures were not significantly different between the brachiocephalic and radiocephlaic fistulas (ρ = 0.871). Percutaneous treatment of thrombosed native arteriovenous

Percutaneous treatment of thrombosed native arteriovenous dialysis fistula insufficiency: efficacy of mechanical thrombectomy with using the stone basket

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Hwan; Ko, Sung Min; Kim, Mi Jung; Kwon, Jung Hyeok; Sohn, Cheol Ho; Choi, Jin Soo; Park, Kyung Sik [Dongsan Medical Center, Keimyung University College of Medicine, Daegu (Korea, Republic of); Kim, Yong Joo [Andong General Hospital, Andon (Korea, Republic of)

2006-06-15

We wanted to evaluate the procedural success after percutaneous treatment of thrombosed native arteriovenous dialysis fistula insufficiency and the efficacy of performing mechanical thrombectomy with using the stone basket. From March 2004 to June 2005, 36 thrombosed native hemodialysis access shunts in the upper limbs (brachiocephalic fistulas: 16 and radiocephalic fistulas: 20) were percutaneously treated in 30 patients. Declotting procedures were performed with using urokinase (100,00-200,000 unit) and manual catheter-directed thrombo-aspiration in all the patients. Angioplasty (6 mm in diameter and 4 cm in length) was performed at the identified area of the stenosis and /or with maceration of the thrombus. In 14 cases with massive thrombosis that was refractory to the above mentioned declotting procedures, mechanical thrombectomy with using a Wittich nitinol stone basket (Cook, Bloomington, IN) was performed. Data regarding the procedural success rate and the patency rate were analyzed by means of Fischer's exact test, and the Kaplan-Meier method with the Log-rank test was used for statistical inter-group comparisons between the brachiocephalic and radiocephalic fistulas. Successful declotting and restoration of thrill were achieved in 30 of 36 procedures (83%). Reestablishment of normal dialysis for at least one session was achieved in 29 of 36 procedures (81%). The procedural success rate for the brachiocephalic fistulas was 94% compared with 70% for the radiocephalic fistulas, but the difference was not statistically significant ({rho} = 0.104). In the cases with performing mechanical thrombectomy and using the stone basket, procedural success was achieved in 93% (13/14). The expected patency rates at 3, 6 and 12 months were 78%, 61% and 51%, respectively. The patency rates after declotting procedures were not significantly different between the brachiocephalic and radiocephlaic fistulas ({rho} = 0.871). Percutaneous treatment of thrombosed native

Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

Energy Technology Data Exchange (ETDEWEB)

Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-03-01

On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

Organic saffron position in the future household consumption basket and effective structures (Case study

Directory of Open Access Journals (Sweden)

Ahmadreza Shahpouri

2014-09-01

Full Text Available Considering the importance of planning at the production level, it is necessary to identify the consumer's future behavior towards organic products. Thus, using seemingly unrelated regression and data of 2012 for 200 households in mashhad, in the present study we have tried to determine the future share of these products in the household's basket and also survey the effective factors on this share in organic saffron. The results showed that households in mashhad tend to allocate an average of 34 percent of their future consumption basket for organic saffron. The regression results illustrated that while children below 10 years old in the households, willingness to pay the price differential compared with non-organic products and organic labeling positively affected the willingness to pay for organic products, environmental concerns variables created by non-organic products had a negative effect on their willingness. Considering the research results, helping to raise people's awareness about food through different educational and advertising approaches, providing a mechanism to label organic products and assessing the strengths and weaknesses of producing and distributing the organic products are provided as suggestions.

In vitro release of ketoprofen suppositories using the USP basket and the flow-through cell dissolution methods.

Science.gov (United States)

Medina, José Raúl; Padilla, Adrián Roberto; Hurtado, Marcela; Cortés, Alma Rosa; Domínguez-Ramírez, Adriana Miriam

2014-05-01

In order to study the release characteristics of ketoprofen suppositories under the hydrodynamic environment generated by USP Apparatus 1 and 4, the dissolution profiles of the Mexican reference product (100 mg) were determined. Phosphate buffer pH 8 and 1% sodium lauryl sulfate (SLS) aqueous solutions were proved as dissolution mediums. Baskets were rotated at 100 rpm with USP Apparatus 1 and different flow rates from 16-32 mL/min with USP Apparatus 4 were used. Drug samples were taken and quantified during 60 min by UV analysis at 260 nm. Mean dissolution time (MDT) and dissolution efficiency (DE) were calculated by model-independent methods. Data were also fitted to several kinetic models. Poor dissolution was found in both dissolution mediums when USP basket method was used ( 0.99). The results suggest the need to establish an adequate dissolution method to evaluate the release kinetics of ketoprofen from suppositories.

Novel method for the measurement of liquid film thickness during fuel spray impingement on surfaces.

Science.gov (United States)

Henkel, S; Beyrau, F; Hardalupas, Y; Taylor, A M K P

2016-02-08

This paper describes the development and application of a novel optical technique for the measurement of liquid film thickness formed on surfaces during the impingement of automotive fuel sprays. The technique makes use of the change of the light scattering characteristics of a metal surface with known roughness, when liquid is deposited. Important advantages of the technique over previously established methods are the ability to measure the time-dependent spatial distribution of the liquid film without a need to add a fluorescent tracer to the liquid, while the measurement principle is not influenced by changes of the pressure and temperature of the liquid or the surrounding gas phase. Also, there is no need for non-fluorescing surrogate fuels. However, an in situ calibration of the dependence of signal intensity on liquid film thickness is required. The developed method can be applied to measure the time-dependent and two-dimensional distribution of the liquid fuel film thickness on the piston or the liner of gasoline direct injection (GDI) engines. The applicability of this technique was evaluated with impinging sprays of several linear alkanes and alcohols with different thermo-physical properties. The surface temperature of the impingement plate was controlled to simulate the range of piston surface temperatures inside a GDI engine. Two sets of liquid film thickness measurements were obtained. During the first set, the surface temperature of the plate was kept constant, while the spray of different fuels interacted with the surface. In the second set, the plate temperature was adjusted to match the boiling temperature of each fuel. In this way, the influence of the surface temperature on the liquid film created by the spray of different fuels and their evaporation characteristics could be demonstrated.

14 CFR Appendix A to Part 31 - Instructions for Continued Airworthiness

Science.gov (United States)

2010-01-01

... that covers details regarding servicing of balloon components, including burner nozzles, fuel tanks... envelope, controls, rigging, basket structure, fuel systems, instruments, and heater assembly that provides... preparation including any storage limits. (i) Instructions for repair on the balloon envelope and its basket...

Assembling Appliances Standards from a Basket of Functions

Energy Technology Data Exchange (ETDEWEB)

Siderious, Hans-Paul; Meier, Alan

2014-08-11

Rapid innovation in product design challenges the current methodology for setting standards and labels, especially for electronics, software and networking. Major problems include defining the product, measuring its energy consumption, and choosing the appropriate metric and level for the standard. Most governments have tried to solve these problems by defining ever more specific product subcategories, along with their corresponding test methods and metrics. An alternative approach would treat each energy-using product as something that delivers a basket of functions. Then separate standards would be constructed for the individual functions that can be defined, tested, and evaluated. Case studies of thermostats, displays and network equipment are presented to illustrate the problems with the classical approach for setting standards and indicate the merits and drawbacks of the alternative. The functional approach appears best suited to products whose primary purpose is processing information and that have multiple functions.

EpiBasket: how e-commerce tools can improve epidemiological preparedness.

Science.gov (United States)

Xing, Weijia; Hejblum, Gilles; Valleron, Alain-Jacques

2013-10-31

Should an emerging infectious disease outbreak or an environmental disaster occur, the collection of epidemiological data must start as soon as possible after the event's onset. Questionnaires are usually built de novo for each event, resulting in substantially delayed epidemiological responses that are detrimental to the understanding and control of the event considered. Moreover, the public health and/or academic institution databases constructed with responses to different questionnaires are usually difficult to merge, impairing necessary collaborations. We aimed to show that e-commerce concepts and software tools can be readily adapted to enable rapid collection of data after an infectious disease outbreak or environmental disaster. Here, the 'customers' are the epidemiologists, who fill their shopping 'baskets' with standardised questions. For each epidemiological field, a catalogue of questions is constituted by identifying the relevant variables based on a review of the published literature on similar circumstances. Each question is tagged with information on its source papers. Epidemiologists can then tailor their own questionnaires by choosing appropriate questions from this catalogue. The software immediately provides them with ready-to-use forms and online questionnaires. All databases constituted by the different EpiBasket users are interoperable, because the corresponding questionnaires are derived from the same corpus of questions. A proof-of-concept prototype was developed for Knowledge, Attitudes and Practice (KAP) surveys, which is one of the fields of the epidemiological investigation frequently explored during, or after, an outbreak or environmental disaster. The catalogue of questions was initiated from a review of the KAP studies conducted during or after the 2003 severe acute respiratory syndrome epidemic. Rapid collection of standardised data after an outbreak or environmental disaster can be facilitated by transposing the e-commerce paradigm

A model of the dose rate calculation for a spent fuel storage structure by Monte Carlo method using the modulated code system SCALE 4.4a

International Nuclear Information System (INIS)

Pantazi, D.; Mateescu, S.; Stanciu, M.; Mete, M.

2001-01-01

The modulated code system SCALE is used to perform a standardized shielding analysis for any facility containing spent fuel: handling devices, transport cask, intermediate and final storage facility. The neutron and gamma sources as well as the dose rates can be obtained using either discrete-ordinates or Monte Carlo methods. The shielding analysis control modules (SAS1, SAS2H and SAS4) provide a general procedure for cross-section preparation, fuel depletion/decay calculation and general onedimensional or multi-dimensional shielding analysis. The module SAS4 used in the analysis presented in this paper, is a three-dimensional Monte Carlo shielding analysis module, which uses an automated biasing procedure specialized for a nuclear fuel transport or storage container. The Spent Fuel Interim Storage Facility in our country is projected to be a parallelepiped concrete monolithic module, consisting of an external reinforced concrete structure with vertical storage cylinders (pits) arranged in a rectangular array. A pit is filled with sealed cylindrical baskets of stainless steel arranged in a stack, and with each basket containing spent fuel bundles in vertical position. The pit is closed with a concrete plug. The cylindrical geometry model is used in the shielding evaluation for a spent fuel storage structure (pit), and only the active parts of the superposed bundles is considered. The dose rates have been calculated in both the axial and radial directions using SAS4.(author)

Regulation of RNA-binding proteins affinity to export receptors enables the nuclear basket proteins to distinguish and retain aberrant mRNAs.

Science.gov (United States)

Soheilypour, M; Mofrad, M R K

2016-11-02

Export of messenger ribonucleic acids (mRNAs) into the cytoplasm is a fundamental step in gene regulation processes, which is meticulously quality controlled by highly efficient mechanisms in eukaryotic cells. Yet, it remains unclear how the aberrant mRNAs are recognized and retained inside the nucleus. Using a new modelling approach for complex systems, namely the agent-based modelling (ABM) approach, we develop a minimal model of the mRNA quality control (QC) mechanism. Our results demonstrate that regulation of the affinity of RNA-binding proteins (RBPs) to export receptors along with the weak interaction between the nuclear basket protein (Mlp1 or Tpr) and RBPs are the minimum requirements to distinguish and retain aberrant mRNAs. Our results show that the affinity between Tpr and RBPs is optimized to maximize the retention of aberrant mRNAs. In addition, we demonstrate how the length of mRNA affects the QC process. Since longer mRNAs spend more time in the nuclear basket to form a compact conformation and initiate their export, nuclear basket proteins could more easily capture and retain them inside the nucleus.

Handling tool for, and method of use of, ice basket removable cruciform

International Nuclear Information System (INIS)

Scrabis, C.M.; Mazza, G.E.; Golick, L.R.; Pomaibo, P.

1987-01-01

This patent describes a method for selectively installing and removing a removable cruciform of a resiliently compressible type, in and from the interior of a generally cylindrical, elongated and vertically oriented ice containment structure, which interior is accessible from an open, upper and thereof. Each such removable cruciform comprises a central housing defining an axis of compression and expansion, means normally biasing the housing in the direction of expansion, legs extending from the housing in a generally radially oriented direction and means on the outer edges of the legs configured to engage a corresponding stiffenning ring rigidly secured on the interior of the cylindrical sidewall of the ice basket, comprising: providing a handling tool having an elongated frame defining an axis, guide means connected to the frame and comprising at least first and second guide fingers extending radially from the axis and angularly displaced thereabout and selectively moveable between retracted and extended positions, and clamping means mounted to the frame and extending axially outwardly therefrom, positioning the tool, with the clamping shoes in the disengaged positions, in axial alignment with and disposed radially outwardly about the central housing of a removable cruciform of the resiliently compressible type; selectively actuating the clamping means to the engaged positions of the clamping shoes for engaging and compressing the central housing of the cruciform and retracting the legs thereof; lowering the tool with the cruciform engaged thereby, axially downwardly through the ice basket to a desired elevation

Development of metal cask for nuclear spent fuel

International Nuclear Information System (INIS)

Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

2001-01-01

It is one of the realistic solutions against increasing demand on interim storage of spent fuel assemblies arising from nuclear power plants in Japan to apply dual purpose (transport and storage) metal casks. Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities, etc. in Japan, and have established transport and storage cask design ''MSF series'' which realizes higher payload and reliability for long term storage. MSF series transport and storage casks use various new design concepts and materials to improve thermal performance of the cask, structural integrity of the basket, durability of the neutron shielding material and so on. This paper summarizes an outline of the cask design that can accommodate BWR spent fuel assemblies as well as the new technologies applied to the design and fabrication. (author)

100KE/KW fuel storage basin surface volumetric factors

International Nuclear Information System (INIS)

Conn, K.R.

1996-01-01

This Supporting Document presents calculations of surface Volumetric factors for the 100KE and 100KW Fuel Storage Basins. These factors relate water level changes to basin loss or additions of water, or the equivalent water displacement volumes of objects added to or removed from the basin

Surface modification of a proton exchange membrane and hydrogen storage in a metal hydride for fuel cells

Science.gov (United States)

Andrews, Lisa

Interest in fuel cell technology is rising as a result of the need for more affordable and available fuel sources. Proton exchange membrane fuel cells involve the catalysis of a fuel to release protons and electrons. It requires the use of a polymer electrolyte membrane to transfer protons through the cell, while the electrons pass through an external circuit, producing electricity. The surface modification of the polymer, NafionRTM, commonly researched as a proton exchange membrane, may improve efficiency of a fuel cell. Surface modification can change the chemistry of the surface of a polymer while maintaining bulk properties. Plasma modification techniques such as microwave discharge of an argon and oxygen gas mixture as well as vacuum-ultraviolet (VUV) photolysis may cause favorable chemical and physical changes on the surface of Nafion for improved fuel cell function. A possible increase in hydrophilicity as a result of microwave discharge experiments may increase proton conductivity. Grafting of acrylic acid from the surface of modified Nafion may decrease the permeation of methanol in a direct methanol fuel cell, a process which can decrease efficiency. Modification of the surface of Nafion samples were carried out using: 1) An indirect Ar/O2 gas mixture plasma investigating the reaction of oxygen radicals with the surface, 2) A direct Ar/O2 gas mixture plasma investigating the reaction of oxygen radicals and VUV radiation with the surface and, 3) VUV photolysis investigating exclusively the interaction of VUV radiation with the surface and any possible oxidation upon exposure to air. Acrylic acid was grafted from the VUV photolysed Nafion samples. All treated surfaces were analyzed using X-ray photoelectron spectroscopy (XPS). Fourier transform infrared spectroscopy (FTIR) was used to analyze the grafted Nafion samples. Scanning electron microscopy (SEM) and contact angle measurements were used to analyze experiments 2 and 3. Using hydrogen as fuel is a

Spent Nuclear Fuel (SNF) Project Product Specification

International Nuclear Information System (INIS)

PAJUNEN, A.L.

2000-01-01

The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

Spent Nuclear Fuel (SNF) Project Product Specification

Energy Technology Data Exchange (ETDEWEB)

PAJUNEN, A.L.

2000-12-07

The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

International Nuclear Information System (INIS)

Sabourin, G.

1998-01-01

This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

[Spatial pattern of land surface dead combustible fuel load in Huzhong forest area in Great Xing'an Mountains].

Science.gov (United States)

Liu, Zhi-Hua; Chang, Yu; Chen, Hong-Wei; Zhou, Rui; Jing, Guo-Zhi; Zhang, Hong-Xin; Zhang, Chang-Meng

2008-03-01

By using geo-statistics and based on time-lag classification standard, a comparative study was made on the land surface dead combustible fuels in Huzhong forest area in Great Xing'an Mountains. The results indicated that the first level land surface dead combustible fuel, i. e., 1 h time-lag dead fuel, presented stronger spatial auto-correlation, with an average of 762.35 g x m(-2) and contributing to 55.54% of the total load. Its determining factors were species composition and stand age. The second and third levels land surface dead combustible fuel, i. e., 10 h and 100 h time-lag dead fuels, had a sum of 610.26 g x m(-2), and presented weaker spatial auto-correlation than 1 h time-lag dead fuel. Their determining factor was the disturbance history of forest stand. The complexity and heterogeneity of the factors determining the quality and quantity of forest land surface dead combustible fuels were the main reasons for the relatively inaccurate interpolation. However, the utilization of field survey data coupled with geo-statistics could easily and accurately interpolate the spatial pattern of forest land surface dead combustible fuel loads, and indirectly provide a practical basis for forest management.

3D modeling of missing pellet surface defects in BWR fuel

Energy Technology Data Exchange (ETDEWEB)

Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov; Williamson, R.L.; Stafford, D.S.; Novascone, S.R.; Hales, J.D.; Pastore, G.

2016-10-15

Highlights: • A global/local analysis procedure for missing pellet surface defects is proposed. • This is applied to defective BWR fuel under blade withdrawal and high power ramp conditions. • Sensitivity of the cladding response to key model parameters is studied. - Abstract: One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can be used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed here. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding

Cross-correlations between Renminbi and four major currencies in the Renminbi currency basket

Science.gov (United States)

Wang, Gang-Jin; Xie, Chi

2013-03-01

We investigate the cross-correlations between Renminbi (CNY) and four major currencies (USD, EUR, JPY, and KRW) in the Renminbi currency basket, i.e., the cross-correlations of CNY-USD, CNY-EUR, CNY-JPY, and CNY-KRW. Qualitatively, using a statistical test in analogy to the Ljung-Box test, we find that cross-correlations significantly exist in CNY-USD, CNY-EUR, CNY-JPY, and CNY-KRW. Quantitatively, employing the detrended cross-correlation analysis (DCCA) method, we find that the cross-correlations of CNY-USD, CNY-EUR, CNY-JPY, and CNY-KRW are weakly persistent. We use the DCCA cross-correlation coefficient ρ to quantify the level of cross-correlations and find the currency weight in the Renminbi currency basket is arranged in the order of USD>EUR>JPY >KRW. Using the method of rolling windows, which can capture the time-varying cross-correlation scaling exponents, we find that: (i) CNY and USD are positively cross-correlated over time, but the cross-correlations of CNY-USD are anti-persistent during the US sub-prime crisis and the European debt crisis. (ii) The cross-correlation scaling exponents of CNY-EUR have the cyclical fluctuation with a nearly two-year cycle. (iii) CNY-JPY has long-term negative cross-correlations, during the European debt crisis, but CNY and KRW are positively cross-correlated.

Impact of Aluminum on Anticipated Corrosion in a Flooded spent nuclear fuel Multi -Canister Overpack

International Nuclear Information System (INIS)

DUNCAN, D.R.

1999-01-01

Corrosion reactions in a flooded MCO are examined to determine the impact of aluminum corrosion products (from aluminum basket grids and spacers) on bound water estimates and subsequent fuel/environment reactions during storage. The mass and impact of corrosion products were determined to be insignificant, validating the choice of aluminum as an MCO component and confirming expectations that no changes to the Technical Databook or particulate mass or water content are necessary

Development of Vegetation and Surface Fuels Following Fire Hazard Reduction Treatment

Directory of Open Access Journals (Sweden)

Scott L. Stephens

2012-08-01

Full Text Available In dry western Unites States forests where past resource management has altered the ecological role of fire and stand characteristics alike, mechanical thinning and prescribed burning are commonly applied in wildfire hazard abatement. The reduced surface fuel loads and stand structures resulting from fuels modifications are temporary, yet few studies have assessed the lifespan of treatment effects. We sampled forest fuels and vegetation following fuels reduction in a chronosequence of time since treatment in the northern Sierra Nevada and southern Cascade regions of California. Treatments altered overstory characteristics including stand density, basal area, and species composition. These effects were still present on the oldest treatment sites (8–15 years post-treatment. Other stand characteristics, particularly timelag fuel loads, seedling density, and shrub cover, exhibited substantial variability, and differences between treatment age classes and between treatment and control groups were not statistically significant.

Used fuel rail shock and vibration testing options analysis

Energy Technology Data Exchange (ETDEWEB)

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

2014-09-25

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

Prototypical fabrication of PWR spent fuel shipping cask

International Nuclear Information System (INIS)

Kwack, Eun Ho; Kim, Byung Ku; Kang, Hee Yung; Lee, Chung Young; Jeon, Kyeong Lak; Lee, Bum Soo

1985-02-01

This report describes about the safety analysis for the spent fuel shipping cask, which is used to transfer a single fuel assembly discharged from PWR in operation in Korea. The contents cover the methods and the results of structural, thermal, thermo-hydraulic, radiation shield and criticality detail analysis. The safety evaluation has been made under the normal transportation and hypothetical accident conditions such as 30ft free drop, puncture, fire, immersion, penetration, corner drop, etc,. Some corrections in design are made, and a brief information for fabrication and transportation are obtained by the use of a 1/6 scale model. The design is based on one year cooling time of the spent fuel with 40,000 MWT/MTU maximum burnup, which gives 7.2KW decay heat and 1.6x10 6 ci/hr radiation intensity. The cask is composed of main body with the double closures, impact limiter and fuel basket. The inner shell, inner closure and valves constitute the pressure boundary of the containment. The inner, intermediate and outer shells, upper and lower forgings are made of stainless steel which compose the main body with lead for gamma shield and 50% ethylene glycol for neutron shield. The impact limiters are made of balsa wood on both end sides of the cask to protect the cask from a sudden shocks in accident during the transportation. The analysis results show that the cask is proved to retain its structural integrity within allowable stress and to be safe under the normal and hypothetical accident conditions, and the maximum dose rates of radiation at 2m distance from the surface of the cask are less than the required values. The weight will be 23.2tons in dry and 27.8 tons in wet with fuel loaded. All the design data, calculated results for the structural integrity, shield and thermal analysis are shown in this report with the basic drawings. (Author)

Safe transport of spent fuels after long-term storage

International Nuclear Information System (INIS)

Aritomi, M.; Takeda, T.; Ozaki, S.

2004-01-01

Considering the scarcity of energy resources in Japan, a nuclear energy policy pertaining to the spent fuel storage has been adopted. The nuclear energy policy sets the rules that spent fuels generated from LWRs shall be reprocessed and that plutonium and unburnt uranium shall be recovered and reused. For this purpose, a reprocessing plant, which has a reprocessing capability of 800 ton/yr, is under construction at Rokkasho Village. However, it is anticipated that the start of its operation will be delayed. In addition, the amount of spent fuels generated from nuclear power plants exceeds its reprocessing capability. Therefore, the establishment of storage technology for spent fuels becomes an urgent problem in Japan in order to continue smoothly the LWR operations. In this paper, the background of nuclear power generation in Japan is introduced at first. Next, the policy of spent fuel storage in Japan and circumstances surrounding the spent fuels in Japan are mentioned. Furthermore, the major subjects for discussions to settle and improve 'Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facility' in Atomic Energy Society of Japan are discussed, such as the integrity of fuel cladding, basket, shielding material and metal gasket for the long term storage for achieving safe transport of spent fuels after the storage. Finally, solutions to the unsolved subject in establishing the spent fuel interim storage technologies ase introduced accordingly

Optimization in the nuclear fuel cycle II: Surface contamination

International Nuclear Information System (INIS)

Pereira, W.S.; Silva, A.X.; Lopes, J.M.; Carmo, A.S.; Fernandes, T.S.; Mello, C.R.; Kelecom, A.

2017-01-01

Optimization is one of the bases of radioprotection and aims to move doses away from the dose limit that is the borderline of acceptable radiological risk. This work aims to use the monitoring of surface contamination as a tool of the optimization process. 53 surface contamination points were analyzed at a nuclear fuel cycle facility. Three sampling points were identified with monthly mean values of contamination higher than 1 Bq ∙ cm -2 , points 28, 42 and 47. These points were indicated for the beginning of the optimization process

Battery and Fuel Cell Development Goals for the Lunar Surface and Lander

Science.gov (United States)

Mercer, Carolyn R.

2008-01-01

NASA is planning a return to the moon and requires advances in energy storage technology for its planned lunar lander and lunar outpost. This presentation describes NASA s overall mission goals and technical goals for batteries and fuel cells to support the mission. Goals are given for secondary batteries for the lander s ascent stage and suits for extravehicular activity on the lunar surface, and for fuel cells for the lander s descent stage and regenerative fuel cells for outpost power. An overall approach to meeting these goals is also presented.

Performance of fire behavior fuel models developed for the Rothermel Surface Fire Spread Model

Science.gov (United States)

Robert Ziel; W. Matt Jolly

2009-01-01

In 2005, 40 new fire behavior fuel models were published for use with the Rothermel Surface Fire Spread Model. These new models are intended to augment the original 13 developed in 1972 and 1976. As a compiled set of quantitative fuel descriptions that serve as input to the Rothermel model, the selected fire behavior fuel model has always been critical to the resulting...

Surface coating Zr or Zr alloy nuclear fuel elements

International Nuclear Information System (INIS)

Donaghy, R.E.; Sherman, A.H.

1980-01-01

A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer. (author)

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

Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

Energy Technology Data Exchange (ETDEWEB)

Kjellstroem, B; Larsson, A E

1967-04-15

In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples.

Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

International Nuclear Information System (INIS)

Kjellstroem, B.; Larsson, A.E.

1967-04-01

In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples

Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

Energy Technology Data Exchange (ETDEWEB)

Sabourin, G. [Atomic Energy of Canada Limited, Montreal, PQ (Canada)

1998-07-01

This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

Near-surface alloys for hydrogen fuel cell applications

DEFF Research Database (Denmark)

Greeley, Jeffrey Philip; Mavrikakis, Manos

2006-01-01

of CO with relatively facile H-2 activation is nearly ideal for this application. We suggest that. as nanoscale materials synthesis techniques improve, it will become feasible to reproducibly prepare NSAs with highly specified surface structures, resulting in the design and manufacture of a wide variety...... facile H-2 activation. These NSAs could, potentially, facilitate highly selective hydrogenation reactions at low temperatures. In the present work, the suitability of NSAs for use as hydrogen fuel cell anodes has been evaluated: the combination of properties, possessed by selected NSAs, of weak binding...... of such materials for use in fuel cells and in an ever. increasing range of catalytic applications. Furthermore, we introduce a new concept for NSA-defect sites, which could be responsible for the promotional catalytic effects of a second metal added. even in minute quantities, to a host metal catalyst....

Mathematical modeling of synthesis gas fueled electrochemistry and transport including H2/CO co-oxidation and surface diffusion in solid oxide fuel cell

Science.gov (United States)

Bao, Cheng; Jiang, Zeyi; Zhang, Xinxin

2015-10-01

Fuel flexibility is a significant advantage of solid oxide fuel cell (SOFC). A comprehensive macroscopic framework is proposed for synthesis gas (syngas) fueled electrochemistry and transport in SOFC anode with two main novelties, i.e. analytical H2/CO electrochemical co-oxidation, and correction of gas species concentration at triple phase boundary considering competitive absorption and surface diffusion. Staring from analytical approximation of the decoupled charge and mass transfer, we present analytical solutions of two defined variables, i.e. hydrogen current fraction and enhancement factor. Giving explicit answer (rather than case-by-case numerical calculation) on how many percent of the current output contributed by H2 or CO and on how great the water gas shift reaction plays role on, this approach establishes at the first time an adaptive superposition mechanism of H2-fuel and CO-fuel electrochemistry for syngas fuel. Based on the diffusion equivalent circuit model, assuming series-connected resistances of surface diffusion and bulk diffusion, the model predicts well at high fuel utilization by keeping fixed porosity/tortuosity ratio. The model has been validated by experimental polarization behaviors in a wide range of operation on a button cell for H2-H2O-CO-CO2-N2 fuel systems. The framework could be helpful to narrow the gap between macro-scale and meso-scale SOFC modeling.

Multi-institutional Evaluation of Upper Urinary Tract Biopsy Using Backloaded Cup Biopsy Forceps, a Nitinol Basket, and Standard Cup Biopsy Forceps.

Science.gov (United States)

Lama, Daniel J; Safiullah, Shoaib; Patel, Roshan M; Lee, Thomas K; Balani, Jyoti P; Zhang, Lishi; Okhunov, Zhamshid; Margulis, Vitaly; Savage, Stephen J; Uchio, Edward; Landman, Jaime

2018-04-06

To compare the performance of 3 contemporary ureteroscopic biopsy devices for the histopathologic diagnosis of upper tract urothelial carcinoma (UTUC). We retrospectively reviewed 145 patients who underwent 182 urothelial biopsies using 2.4F backloaded cup biopsy forceps, a nitinol basket, or 3F standard cup biopsy forceps at 3 tertiary academic centers between 2011 and 2016. Experienced genitourinary pathologists provided an assessment of each specimen without knowledge of the device used for biopsy. For patients who underwent nephroureterectomy without neoadjuvant chemotherapy within 3 months of biopsy-proven UTUC diagnosis, the biopsy grade was compared with both the grade and stage of the surgical specimen. Biopsy utilization varied among the 3 institutions (P cup forceps was rated similarly to the nitinol basket (P >.05) and was favored over standard cup forceps specimens. Grade concordance was not affected by specimen size (P >.05), morphology (P >.1), or location (P >.5). No difference existed among the devices in the rate of acquiring a grade concordant biopsy; however, the backloaded cup forceps provided concordant biopsies that could be distinguished as low- and high-grade (P = .02). The backloaded cup forceps and nitinol basket obtained a higher quality urothelial specimen compared with standard cup forceps. Ureteroscopic biopsy device selection did not significantly impact the accuracy of the histologic diagnosis of UTUC. Copyright © 2018 Elsevier Inc. All rights reserved.

Digital image processing: Cylindrical surface plane development of CAREM fuel pellets

International Nuclear Information System (INIS)

Caccavelli, J; Cativa Tolosa, S; Gommes, C

2012-01-01

As part of the development of fuel pellets (FPs) for nuclear reactor CAREM-25, is necessary to systematize the analysis of the mechanical integrity of the FPs that is now done manually by a human operator. Following specifications and standards of reference for this purpose, the FPs should be inspected visually for detecting material discontinuities in the FPs surfaces to minimize any deterioration, loss of material and excessive breakage during operation and load of fuel bars. The material discontinuities are classified into two defects: surface cracks and chips. For each of these surface defects exist acceptance criteria that determine if the fuel pellet (FP) as a whole is accepted or rejected. One criteria for surface cracks is that they do not exceed one third (1/3) of the circumferential surface of the FP. The FP has cylindrical shape, so some of these acceptance criteria make difficult to analyze the FP in a single photographic image. Depending on the axial rotation of the FP, the crack could not be entirely visualized on the picture frame. Even a single crack that appears in different parts of the FP rotated images may appear to be different cracks in the FP when it is actually one. For this reason it is necessary, for the automatic detection and measurement of surface defects, obtain the circumferential surface of the FP into a single image in order to decide the acceptance or reject of the FP. As the FP shape is cylindrical, it is possible to obtain the flat development of the cylindrical surface (surface unrolling) of the FPs into a single image combining the image set of the axial rotation of the FP. In this work, we expose the procedure to implement the flat development of the cylindrical surface (surface unrolling). Starting from a photographic image of the FP surface, which represents the projection of a cylinder in the plane, we obtain three-dimensional information of each point on the cylindrical surface of the FP (3D-mapping). Then, we can

CLUSTER ANALYSIS UNTUK MEMPREDIKSI TALENTA PEMAIN BASKET MENGGUNAKAN JARINGAN SARAF TIRUAN SELF ORGANIZING MAPS (SOM

Directory of Open Access Journals (Sweden)

Gregorius Satia Budhi

2008-01-01

Full Text Available Basketball World has grown rapidly as the time goes on. This is signed by many competition and game all over the world. With the result there are many basketball players with their different playing characteristics. Demand for a coach or scout to look for or search great players to make a solid team as a coach requirement. With this application, a coach or scout will be helped in analyzing in decision making. This application uses Self Organizing Maps algorithm (SOM for Cluster Analysis. The real NBA player data is used for competitive learning or training process and real player data from Indonesian or Petra Christian University Basketball Players is used for testing process. The NBA Player data is prepared through cleaning process and then is transformed into a form that can be processed by SOM Algorithm. After that, the data is clustered with the SOM algorithm. The result of that clusters is displayed into a form that is easy to view and analyze. This result can be saved into a text file. By using the output / result of this application, that are the clusters of NBA player, the user can see the statistics of each cluster. With these cluster statistics coach or scout can predict the statistic and the position of a testing player who is in the same cluster. This information can give a support for the coach or scout to make a decision. Abstract in Bahasa Indonesia : Dunia bola basket telah berkembang dengan pesat seiring dengan berjalannya waktu. Hal ini ditandai dengan munculnya berbagai macam dan jenis kompetisi dan pertandingan baik dunia maupun dalam negeri. Sehingga makin banyak dilahirkannya pemain berbakat dengan berbagai karakteristik permainan yang berbeda. Tuntutan bagi seorang pelatih/pemandu bakat, untuk dapat melihat secara jeli dalam memenuhi kebutuhan tim untuk membentuk tim yang solid. Dengan dibuatnya aplikasi ini, maka akan membantu proses analisis dan pengambilan keputusan bagi pelatih maupun pemandu bakat Aplikasi ini

Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

Science.gov (United States)

Najeeb, Umair

This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

Food group contribution of essential elements of the Sao Paulo State market basket

International Nuclear Information System (INIS)

Avegliano, Roseane P.; Maihara, Vera A.; Silva, Fabio F. da

2009-01-01

To establish a Market Basket of Sao Paulo state seventy-one foods, with a mean consumption of more than 2 g day -1 per person, were grouped into 30 food categories. The food groups were: cereals, leguminous, leafy vegetables, fruity vegetables, tuberous vegetables, tropical fruits, other fruits, flours, pastas, breads, biscuits, prime grade beef, standard grade beef, pork meats, other meats, poultry, milk/cream, other dairy products, sugars, sweets, salts, sauces, oils, fats, alcoholic beverages, non-alcoholic beverages, coffee, ready-made dishes, saltwater and freshwater fishes. Information about individual food consumption was obtained from a recent national household food budget survey 'POF 2002-2003' conducted by the Brazilian Institute for Geography and Statistics from July 2002 to June 2003. Sampling and kitchen preparation of foods were carried out in restaurants of the University of Sao Paulo. Each food item was individually prepared table-ready. Foods of the same group were mixed, homogenized, pulverized and analyzed for the determination of Ca, Cr, Fe, K, Na and Zn concentrations by Instrumental Neutron Activation Analysis. Average daily intake of each element was calculated by multiplying the element concentration in the food by the corresponding weight of the ready-to-consume food group. The contribution of each food group to the total daily intake of elements by the ready-to-consume food groups of the Market Basket was evaluated. The food groups representing the highest contributions were salts: 79% Na; breads: 37% Fe and 46% Cr; cereals: 19% Zn and milk/cream: 58% Ca and 24% K. (author)

Graphite anode surface modification with controlled reduction of specific aryl diazonium salts for improved microbial fuel cells power output.

Science.gov (United States)

Picot, Matthieu; Lapinsonnière, Laure; Rothballer, Michael; Barrière, Frédéric

2011-10-15

Graphite electrodes were modified with reduction of aryl diazonium salts and implemented as anodes in microbial fuel cells. First, reduction of 4-aminophenyl diazonium is considered using increased coulombic charge density from 16.5 to 200 mC/cm(2). This procedure introduced aryl amine functionalities at the surface which are neutral at neutral pH. These electrodes were implemented as anodes in "H" type microbial fuel cells inoculated with waste water, acetate as the substrate and using ferricyanide reduction at the cathode and a 1000 Ω external resistance. When the microbial anode had developed, the performances of the microbial fuel cells were measured under acetate saturation conditions and compared with those of control microbial fuel cells having an unmodified graphite anode. We found that the maximum power density of microbial fuel cell first increased as a function of the extent of modification, reaching an optimum after which it decreased for higher degree of surface modification, becoming even less performing than the control microbial fuel cell. Then, the effect of the introduction of charged groups at the surface was investigated at a low degree of surface modification. It was found that negatively charged groups at the surface (carboxylate) decreased microbial fuel cell power output while the introduction of positively charged groups doubled the power output. Scanning electron microscopy revealed that the microbial anode modified with positively charged groups was covered by a dense and homogeneous biofilm. Fluorescence in situ hybridization analyses showed that this biofilm consisted to a large extent of bacteria from the known electroactive Geobacter genus. In summary, the extent of modification of the anode was found to be critical for the microbial fuel cell performance. The nature of the chemical group introduced at the electrode surface was also found to significantly affect the performance of the microbial fuel cells. The method used for

Thermal evaluation facility for LMFBR spent fuel transport

International Nuclear Information System (INIS)

Wesley, D.A.

1980-04-01

A full-scale mock-up of a 217 pin breeder reactor fuel assembly in a cylindrical pipe was initially designed and constructed by Oak Ridge National Laboratory (ORNL). It was transferred to Sandia where it was extensively redesigned and modified. The 217 pin hexagonal core assembly was installed in a smaller diameter stainless steel pipe which more closely represents the diameter of a shipping canister or shipping cask basket wall. Two-hundred four of the tubes are electrically heated over an active length of 4-feet and the remaining thirteen are instrumented with multiple junction thermocouples which can be traversed axially. Thermocouples and heat-flux gauges are located on the hex core and canister perimeters at several axial locations

Criticality evaluations of scrambled fuel in water basin storage

International Nuclear Information System (INIS)

Fast, E.

1989-01-01

Fuel stored underwater in the Idaho Chemical Processing Plant basins has been subjected to the usual criticality safety evaluations to assure safe storage configurations. Certain accident or emergency conditions, caused by corrosion or a seismic event, could change the fuel configuration and environment to invalidate previous calculations. Consideration is given here to such contingencies for fuel stored in three storage basins. One basin has fuel stored in racks, on a generally flat floor. In the other two basins, the fuel is stored on yokes and in baskets suspended from a monorail system. The floor is ribbed with 30.48-cm-thick and 80-cm-high concrete barriers across the basin width and spaced 30.48 cm apart. The suspended fuel is typically down to 15 cm above the floor of the channel between the concrete barriers. These basins each have 29 channels of 18 positions maximum per channel for a total of 522 possible positions, which are presently 77 and 49% occupied. The three basins are hydraulically interconnected. Several scenarios indicate possible changes in the fuel configuration. An earthquake could rupture a basin wall or floor, allowing the water to drain from all basins. All levels of water would fall to the completely drained condition. Suspended fuel could drop and fall over within the channel. Corrosion might weaken the support systems or cause leaks in sealed fuel canisters. Calculations were made with the KENO-IV criticality program and the library of mostly Hansen-Roach 16-energy-group neutron cross sections

An assessment of the transportation costs of shipping non-fuel assembly hardware to FWMS facilities

International Nuclear Information System (INIS)

Shappert, L.B.; Joy, D.S.; Johnson, P.E.; Danese, F.L.; Best, R.E.

1991-01-01

This study examines the cost of using Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) Initiative I casks for transporting 62,700 MTU of spent fuel plus associated non-fuel assembly hardware (NFAH) between reactor sites and either a monitored retrievable storage (MRS) or a repository facility. The study further considers the benefits of increasing the cell size of the Initiative I BWR cask baskets to accommodate the fuel plus channels (which also would decrease the capacity of the cask to carry BWR fuel without channels) and the use of a commercial, non-spent-fuel cask to carry compacted NFAH that could not be shipped integrally. Costs that are developed involve transportation charges, capital costs for casks, and canning costs of NFAH that have been separated from the fuel. The results indicate that significant cost savings are possible if NFAH is accepted into the Federal Waste Management System (FWMS) that is either integral with the spent fuel, or consolidated if it has been separated. Shipment of unconsolidated NFAH is very expensive. Transportation costs for shipping to a western repository are approximately 50 to 75% higher than the costs for shipping to an eastern MRS

Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

International Nuclear Information System (INIS)

Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai

2002-01-01

Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis

Kajian Algoritma Sequential Pattern Mining Dan Market Basket Analysis Dalam Pengenalan Pola Belanja Customer Untuk Layout Toko

Directory of Open Access Journals (Sweden)

Rusito Rusito

2016-01-01

Full Text Available Penelitian ini membahas tentang keterkaitan antar item yang dibeli oleh customer dalam toko ritel. Pengetahuan keterkaitan item yang dibeli dapat digunakan untuk  menentukan tata letak barang dagangan toko ritel. Hal ini penting agar konsumen dapat mudah mendapatkan barang yang dibutuhkan. Sehingga dapat meningkatkan omzet penjualan toko ritel sehingga akhirnya menambah keuntungan bagi pemilik toko ritel. Teknik yang digunakan untuk menyelesaikan penggalian data dan keterkaitan pembelian tersebut menggunakan pendekatan Association rule dan Market Basket Analysis. Sedangkan untuk mencari keterkaitan item tersebut digunakan algoritma Sequential Pattern Mining. Digunakan karena mampu menangani jumlah database yang besar dan sangat baik disisi kecepatan pemrosesan. Berbagai aplikasi telah diidentifikasi, termasuk misalnya, cross-selling, analisis situs Web, pendukung keputusan, evaluasi kredit, acara prediksi kriminal, analisis perilaku pelanggan  dan deteksi penipuan. Dari penelitian yang telah dilakukan diperoleh  pola-pola belanja customer untuk membentuk suatu layout display dalam toko ritel. Penelitian ini juga menyajikan suatu kerja algoritma yang lebih efektif dari algoritma asli karena terdapat pembatasan perulangan. Untuk kombinasi maksimal 5 item dengan waktu eksekusi 421.06 detik untuk 200 nota.   Kata kunci : Data Mining, Algoritma Sequential Pattern Mining, Market Basket Analysis, Apriori, Layout, Toko Ritel

Surface roughness effect on the metallic bipolar plates of a proton exchange membrane fuel cell

International Nuclear Information System (INIS)

Lin, Chien-Hung

2013-01-01

Highlights: ► Various degrees of roughness are caused by the sandblasting method. ► An improper surface modification depletes the PEMFC performance severely. ► The AC impedance are used to assess the fuel gas transfer effect. ► The Warburg resistance form in the coarse flow channel surface. - Abstract: Proton exchange membrane fuel cells (PEMFCs) is a promising candidate as energy systems. However, the stability and lifetime of cells are still important issues. The effect of surface roughness on metallic bipolar plate is discussed in this paper. Various roughness on the bulk surface are obtained by the sandblasting method. The grain sizes of sand are selected as 50, 100 and 200 μm. The Ac impedance experiment results show that the bipolar plate roughness and carbon paper porosity are well matched when the surface roughness is within 1–2 μm. Superior condition decreases the contact resistance loss in the fuel cell. The high frequency resistance of the coarse surface was larger than that of the substrate by around 5 mΩ. Furthermore, a new arc was formed at the low frequency region. Hence, the unmatch roughness condition of the bipolar plate significantly increases the contact resistance and mass transfer resistance. This paper develops a sequential approach to study an optimum surface roughness by combining the whole performance (I–V) curve and AC impedance result. It benefits us to quantify the contact and mass transfer resistance exists in the PEMFC. The proposed surface treatment improves the surface effect and promotes the implement of potential metallic bipolar plate in near future

Effects of piston surface treatments on performance and emissions of a methanol-fueled, direct injection, stratified charge engine

Energy Technology Data Exchange (ETDEWEB)

West, B.; Green, J.B. [Oak Ridge National Lab., TN (United States)

1994-07-01

The purpose of this study was to investigate the effects of thermal barrier coatings and/or surface treatments on the performance and emissions of a methanol-fueled, direct-injection, stratified-charge (DISC) engine. A Ricardo Hydra Mark III engine was used for this work and in previous experiments at Oak Ridge National Laboratory (ORNL). The primary focus of the study was to examine the effects of various piston insert surface treatments on hydrocarbon (HC) and oxides of nitrogen (NO{sub x}) emissions. Previous studies have shown that engines of this class have a tendency to perform poorly at low loads and have high unburned fuel emissions. A blank aluminum piston was modified to employ removable piston bowl inserts. Four different inserts were tested in the experiment: aluminum, stainless steel with a 1.27-mm (0.050-in.) air gap (to act as a thermal barrier), and two stainless steel/air-gap inserts with coatings. Two stainless steel inserts were dimensionally modified to account for the coating thickness (1.27-mm) and coated identically with partially stabilized zirconia (PSZ). One of the coated inserts then had an additional seal-coat applied. The coated inserts were otherwise identical to the stainless steel/air-gap insert (i.e., they employed the same 1.27-mm air gap). Thermal barrier coatings were employed in an attempt to increase combustion chamber surface temperatures, thereby reducing wall quenching and promoting more complete combustion of the fuel in the quench zone. The seal-coat was applied to the zirconia to reduce the surface porosity; previous research suggested that despite the possibly higher surface temperatures obtainable with a ceramic coating, the high surface area of a plasma-sprayed coating may actually allow fuel to adhere to the surface and increase the unburned fuel emissions and fuel consumption.

Empowering families by engaging and relating Murri way: a grounded theory study of the implementation of the Cape York Baby Basket program.

Science.gov (United States)

McCalman, Janya; Searles, Andrew; Bainbridge, Roxanne; Ham, Rachael; Mein, Jacki; Neville, Johanna; Campbell, Sandra; Tsey, Komla

2015-05-21

Evaluating program outcomes without considering how the program was implemented can cause misunderstandings and inefficiencies when initiating program improvements. In conjunction with a program evaluation, reported elsewhere, this paper theorises the process of implementing an Indigenous Australian maternal and child health program. The Baby Basket program was developed in 2009 for the remote Cape York region and aimed to improve the attendance and engagement of Indigenous women at antenatal and postnatal clinics through providing three baskets of maternal and baby goods and associated health education. Constructivist grounded theory methods were used to generate and analyse data from qualitative interviews and focus groups with Indigenous women who received the baskets, their extended family members, and healthcare workers who delivered them. Data was coded in NVivo with concepts iteratively compared until higher order constructs and their relationships could be modelled to explain the common purpose for participants, the process involved in achieving that purpose, key strategies, conditions and outcomes. Theoretical terms are italicised. Program implementation entailed empowering families through a process of engaging and relating Murri (Queensland Indigenous) way. Key influencing conditions of the social environment were the remoteness of communities, keeping up with demand, families' knowledge, skills and roles and organisational service approaches and capacities. Engaging and relating Murri way occurred through four strategies: connecting through practical support, creating a culturally safe practice, becoming informed and informing others, and linking at the clinic. These strategies resulted in women and families taking responsibility for health through making healthy choices, becoming empowered health consumers and advocating for community changes. The theoretical model was applied to improve and revise Baby Basket program implementation, including

Spatial variability of surface fuels in treated and untreated ponderosa pine forests of the southern Rocky Mountains

Science.gov (United States)

Emma Vakili; Chad M. Hoffman; Robert E. Keane; Wade T. Tinkham; Yvette Dickinson

2016-01-01

There is growing consensus that spatial variability in fuel loading at scales down to 0.5 m may govern fire behaviour and effects. However, there remains a lack of understanding of how fuels vary through space in wildland settings. This study quantifies surface fuel loading and its spatial variability in ponderosa pine sites before and after fuels treatment in the...

Effect of high surface area activated carbon on thermal degradation of jet fuel

Energy Technology Data Exchange (ETDEWEB)

Gergova, K.; Eser, S.; Arumugam, R.; Schobert, H.H. [Pennsylvania State Univ., University Park, PA (United States)

1995-05-01

Different solid carbons added to jet fuel during thermal stressing cause substantial changes in pyrolytic degradation reactions. Activated carbons, especially high surface area activated carbons were found to be very effective in suppressing solid deposition on metal reactor walls during stressing at high temperatures (425 and 450{degrees}C). The high surface area activated carbon PX-21 prevented solid deposition on reactor walls even after 5h at 450{degrees}C. The differences seen in the liquid product composition when activated carbon is added indicated that the carbon surfaces affect the degradation reactions. Thermal stressing experiments were carried out on commercial petroleum-derived JPTS jet fuel. We also used n-octane and n-dodecane as model compounds in order to simplify the study of the chemical changes which take place upon activated carbon addition. In separate experiments, the presence of a hydrogen donor, decalin, together with PX-21 was also studied.

An experiment to examine the mechanistic behaviour of irradiated CANDU fuel stored under dry conditions

International Nuclear Information System (INIS)

Oldaker, I.E.; Crosthwaite, J.L.; Keltie, R.J.; Truss, K.J.

1979-01-01

A program has begun to use the Whiteshell Nuclear Research Establishment dry-storage canisters to store some selected CANDU irradiated fuel bundles in an 'easily retrievable basket.' The object of the experimental program is to study the long-term stability of the Zircaloy-sheathed UO 2 and UC fuel elements when stored in air. Bundles were loaded into a canister in October 1979 following detailed examination and removal of up to three complete elements from most bundles. These elements are currently being subjected to detailed destructive examinations, including metallography and scanning electron micrography, to fully characterize their pre-storage condition. After four years, and every five years thereafter, further elements will be examined similarly to study the effects of the storage environment on the stability of the Zircaloy sheathing, and on its continued ability to contain the fuel safely in an interim storage facility. (author)

Surface-reconstructed graphite nanofibers as a support for cathode catalysts of fuel cells.

Science.gov (United States)

Gan, Lin; Du, Hongda; Li, Baohua; Kang, Feiyu

2011-04-07

Graphite nanofibers (GNFs), on which surface graphite edges were reconstructed into nano-loops, were explored as a cathode catalyst support for fuel cells. The high degree of graphitization, as well as the surface-reconstructed nano-loops that possess topological defects for uniform metal deposition, resulted in an improved performance of the GNF-supported Pt catalyst.

Smoothing the payoff for efficient computation of Basket option prices

KAUST Repository

Bayer, Christian

2017-07-22

We consider the problem of pricing basket options in a multivariate Black–Scholes or Variance-Gamma model. From a numerical point of view, pricing such options corresponds to moderate and high-dimensional numerical integration problems with non-smooth integrands. Due to this lack of regularity, higher order numerical integration techniques may not be directly available, requiring the use of methods like Monte Carlo specifically designed to work for non-regular problems. We propose to use the inherent smoothing property of the density of the underlying in the above models to mollify the payoff function by means of an exact conditional expectation. The resulting conditional expectation is unbiased and yields a smooth integrand, which is amenable to the efficient use of adaptive sparse-grid cubature. Numerical examples indicate that the high-order method may perform orders of magnitude faster than Monte Carlo or Quasi Monte Carlo methods in dimensions up to 35.

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

International Nuclear Information System (INIS)

Andress, D.; McLeod, N.B.; Rahimi, M.; Joy, D.S.; Peterson, R.W.

1991-01-01

The DOE has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design - the GA-4 and GA-9 truck casks and the BR-100 rail cask. The GA-4 cask is designed for PWR fuel only; the GA-9 cask is a longer cask with less shielding designed for BWR fuel only; and the BR-100 cask is designed to accommodate both PWR and BWR fuels. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elements as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. Because of the button and spring interference, the basket openings in these casks will not accommodate assemblies in the BWR/2,3 and BWR/4-6 fuel classes with the fuel channels in place

The dry spent RBMK fuel cask storage site at the Ignalina NPP in Lithuania

International Nuclear Information System (INIS)

Penkov, V.V.; Diersch, R.

1999-01-01

At present, there are about 15,000 spent RBMK fuel assemblies stored in the water pools near the reactors at the Ignalina Nuclear Power Plant (INPP). Part of them are cut in two bundles and stored in standardized baskets in the pools. Each basket is loaded with 102 bundles. For long-term interim storage of this fuel, it was decided to use dry storage in casks. For this reason, the total activity to be stored is split into individual units (casks). Each cask represents a closed and independent safety system, fulfilling all safety-relevant requirements for both normal operational and hypothetical accidental conditions. The main safety relevant features of the storage cask system are: (1) Inherent safety system; (2) Double barrier system; (3) Passive cooling by natural convection; (4) Safety against accidents. The cask dry storage system is a cost effective and multi-functional system for storage, transport after the operation time and final disposal under consideration of additional protective elements. From an economical point of view, cask storage has a number of advantages. Two cask types have been intended for the INPP storage site: (1) The CASTOR RBMK cask made of ductile cast iron; (2) The CONSTOR RBMK sandwich cask made of an inner and outer steel shell and reinforced heavy concrete. The CASTOR RBMK and the CONSTOR RBMK casks are designed to withstand severe storage site accidents and with help of impact limiters - to fulfil the IAEA test criteria for type B(U)F packages. The INPP spent RBMK fuel storage site is designed as an open air storage for an operational time of 50 years. The casks are arranged on the concrete storage pad. The site is equipped with a crane for cask handling and technological buildings and security systems. The safety analyses for fuel and cask handling and for cask handling and for cask technology at the site have been made and accepted by the Lithuanian Competent Authority. (author)

Handbook for inventorying surface fuels and biomass in the Interior West

Science.gov (United States)

James K. Brown; Rick D. Oberheu; Cameron M. Johnston

1982-01-01

Presents comprehensive procedures for inventorying weight per unit area of living and dead surface vegetation, to facilitate estimation of biomass and appraisal of fuels. Provides instructions for conducting fieldwork and calculating estimates of downed woody material, forest floor litter and duff, herbaceous vegetation, shrubs, and small conifers. Procedures produce...

Apparatus for surface treatment of U-Pu carbide fuel samples

International Nuclear Information System (INIS)

Fukushima, Susumu; Arai, Yasuo; Handa, Muneo; Ohmichi, Toshihiko; Shiozawa, Ken-ichi.

1979-05-01

Apparatus has been constructed for treating the surface of U-Pu carbide fuel samples for EPMA. The treatment is to clean off oxide layer on the surface, then coat with an electric-conductive material. The apparatus, safe in handling plutonium, operates as follows. (1) To avoid oxidation of the analyzing surface by oxygen and water in the air, series of cleaning and coating, i.e. ion-etching and ion-coating or ion-etching and vacuum-evaporation is done at the same time in an inert gas atmosphere. (2) Ion-etching is possible on samples embedded in non-electric-conductive and low heat-conductive resin. (3) Since the temperature rise in (2) is negligible, there is no deterioration of the samples. (author)

Comparative release studies on suppositories using the basket, paddle, dialysis tubing and flow-through cell methods I. Acetaminophen in a lipophilic base suppository.

Science.gov (United States)

Hori, Seiichi; Kawada, Tsubasa; Kogure, Sanae; Yabu, Shinako; Mori, Kenji; Akimoto, Masayuki

2017-02-01

The release characteristics of lipophilic suppositories containing acetaminophen (AAP) were examined using four types of dissolution methods: the basket, paddle, dialysis tubing (DT) and flow-through cell (FTC) methods. The suitability of each apparatus for quality control in AAP compounded suppositories was evaluated using statistical procedures. More than 80% of the drug was released over 60 min in all the release methods studied, with the exception of the basket method. Reproducible and faster release was achieved using the paddle method at 100 and 200 rpm, whereas poor release occurred with the basket method. The mean dissolution time (MDT), maximum dissolved quantity of AAP at the end of the sampling time (Q) and dissolution efficiency (DE) were calculated by model-independent methods. The FTC method with a single chamber used in this study was also appreciable for AAP suppositories (Q of 100%, MDT of 71-91 min and DE of 75-80%). The DT apparatus is considered similar to the FTC apparatus from a quality control perspective for judging the release properties of lipophilic base suppositories containing AAP. However, even the single chamber FTC used in this study has potential as an in vitro drug release test for suppositories. The comparative dissolution method is expected to become one of the valuable tools for selecting an adequate dissolution test.

Experimental determination of fuel surface temperature in the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Khang, Ngo Phu; Huy, Ngo Quang; An, Tran Khac; Lam, Pham Van [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Measured fuel surface temperatures, obtained at various locations of the core of the Dalat Nuclear Research Reactor under normal operating conditions, are presented, and some thermal characteristics of the reactor are discussed. (author). 2 refs., 11 figs., 2 tabs.

Sogin enriched uranium extraction (EUREX) plant spent fuel pool cleaning and decontamination utilizing the Smart Safe solution

International Nuclear Information System (INIS)

Denton, M.S.; Gili, M.; Nasta, M.; Quintiliani, R.; Caccia, G.; Botzen, W.; Forrester, K.

2009-01-01

SOGIN's EUREX facility in Italy was developed as a pilot plant functional testing laboratory for spent fuel reprocessing. This facility was operated successfully for many years since 1970 and was eventually shutdown consistent with Italy's suspension of all nuclear operations. At the time of suspension, the EUREX facility still had spent nuclear fuel assemblies in storage from a nearby PWR. Other fuel assemblies from an Italian AGR had remained stored in the spent fuel pool for the 20 years or so waiting for removal and reprocessing abroad. Being Magnox fuel elements, their recovery for the transport produced a huge amount of sludge in the pool. During this time, sediment, dirt, corrosion products, fuel cladding, etc. has collected within the fuel pool as a crud layer dispersed throughout. Most of this crud has accumulated on the horizontal surfaces of the pool and fuel element assemblies, while some remains as a suspended colloidal material. Furthermore many other contaminated metal components, used during the operation years, where still inside the pool. Due to a pool leak discovered in 2006, SOGIN speeded up its pool decommissioning program, making also available the transfer of the spent fuel to a nearby interim repository, with the goal to completely drain the pool in the shortest period of time. In order for SOGIN to successfully transfer the fuel assemblies from their current storage basket locations to the spent fuel transfer cask, the bulk of the crud needed to be removed. This cleanup operation was deemed necessary to minimize the suspension of contamination in the water during underwater handling operations. This would reduce the decontamination efforts on the transfer cask upon removal, once loaded with the spent fuel, and enhance safety by reducing potential underwater visibility issues. The operations were completed in July 2008 with the release to the environment of the pool water, thoroughly purified and without any relevant radiological impact. The

Molten salt electrorefining method

International Nuclear Information System (INIS)

Tanaka, Hiroshi; Nakamura, Hitoshi; Shoji, Yuichi; Matsumaru, Ken-ichi.

1994-01-01

A molten cadmium phase (lower side) and a molten salt phase (upper side) are filled in an electrolytic bath. A basket incorporating spent nuclear fuels is inserted/disposed in the molten cadmium phase. A rotatable solid cathode is inserted/disposed in the molten salt phase. The spent fuels, for example, natural uranium, incorporated in the basket is dissolved in the molten cadmium phase. In this case, the uranium concentration in the molten salt phase is determined as from 0.5 to 20wt%. Then, electrolysis is conducted while setting a stirring power for stirring at least the molten salt phase of from 2.5 x 10 2 to 1 x 10 4 based on a reynolds number. Crystalline nuclei of uranium are precipitated uniformly on the surface of the solid cathode, and they grow into fine dendrites. With such procedures, since short-circuit between the cathode precipitates and the molten cadmium phase (anode) is scarcely caused, to improve the recovering rate of uranium. (I.N.)

Performances of TN {sup registered} 24 E. An AREVA used fuel transport and interim storage cask for the German market

Energy Technology Data Exchange (ETDEWEB)

Brion, Thomas [AREVA TN International, Montigny Le Bretonneux (France)

2013-07-01

Part of the AREVA Group, TN International offers a complete range of transport and interim storage solutions for radioactive materials throughout the entire nuclear fuel cycle. A world leader in its sector, TN International has supported for 50 years the expansion of the nuclear industry, in particular by providing expertise in secure packing systems for the storage of used fuel assemblies. As an answer to EON and EnBW, two German utilities, needs, TN International has designed and manufactured the TN {sup registered} 24E cask, offering the following high level performances: 1. transport and storage over a period of 40 years of up to 21 PWR spent nuclear fuel (SNF), allowing for example to load up to 17 MOX fuel assemblies and 4 UOX SNF. 2. high flexibility in the fuel assemblies loading plans, inducing no general predefined constraints with regards to the MOX or UOX fuel positions in the basket of the cask Safety margin related to radioprotection, thermal and mechanical behaviour of the fuel assemblies can be calculated loading plan per loading plan. (orig.)

Hanford K Basins spent nuclear fuels project update

International Nuclear Information System (INIS)

Hudson, F.G.

1997-01-01

Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed

Hanford K Basins spent nuclear fuels project update

Energy Technology Data Exchange (ETDEWEB)

Hudson, F.G.

1997-10-17

Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed.

Spices as a source of lead exposure: a market-basket survey in Sri Lanka.

Science.gov (United States)

Senanayake, M P; Perera, R; Liyanaarachchi, L A; Dassanayake, M P

2013-12-01

We performed a laboratory analysis of spices sold in Sri Lanka for lead content. Samples of curry powder, chili powder and turmeric powder from seven provinces, collected using the market basket survey method, underwent atomic absorption spectrometry. Blanks and standards were utilised for instrument calibration and measurement accuracy. The results were validated in two different laboratories. All samples were found to have lead levels below the US Food and Drug Administration's action level of 0.5 μg/g. Spices sold in Sri Lanka contain lead concentrations that are low and within the stipulated safety standards.

Heat transfer from the roughened surface of gas cooled fast breeder reactor fuel element

International Nuclear Information System (INIS)

Tang, I.M.

1979-01-01

The temperature distributions and the augmentation of heat transfer performance by artificial roughening of a gas cooled fast breeder reactor (GCFR) fuel rod cladding are studied. Numerical solutions are based on the axisymmetric assumption for a two-dimensional model for one rib pitch of axial distance. The local and axial clad temperature distributions are obtained for both the rectangular and ramp rib roughened surface geometries. The transformation of experimentally measured convective heat transfer coefficients, in terms of Stanton number, into GCFR values is studied. In addition, the heat transfer performance of a GCFR fuel rod cladding roughened surface design is evaluated. Approximate analytical solution for correlating an average Stanton number is also obtained and satisfactorily compared with the corresponding numerical result for a GCFR design. The analytical correlation is useful in assessing roughened surface heat transfer performance in scoping studies and conceptual design

Predicting surface fuel models and fuel metrics using lidar and CIR imagery in a dense mixed conifer forest

Science.gov (United States)

Marek K. Jakubowksi; Qinghua Guo; Brandon Collins; Scott Stephens; Maggi. Kelly

2013-01-01

We compared the ability of several classification and regression algorithms to predict forest stand structure metrics and standard surface fuel models. Our study area spans a dense, topographically complex Sierra Nevada mixed-conifer forest. We used clustering, regression trees, and support vector machine algorithms to analyze high density (average 9 pulses/m

Prescribed burning and mastication effects on surface fuels in southern pine beetle-killed loblolly pine plantations

Science.gov (United States)

Aaron D. Stottlemyer; Thomas A. Waldrop; G. Geoff Wang

2015-01-01

Surface fuels were characterized in loblolly pine (Pinus taeda L.) plantations severely impacted by southern pine beetle (Dendroctonus frontalis Ehrh.) (SPB) outbreaks in the upper South Carolina Piedmont. Prescribed burning and mastication were then tested as fuel reduction treatments in these areas. Prescribed burning reduced...

Perancangan Dan Pembuatan Modul Data Mining Market Basket Analysis Pada Odoo Dengan Studi Kasus Supermarket X

OpenAIRE

Hendratha, Stefani Natalia; Yulia, Yulia; Budhi, Gregorius Satia

2016-01-01

Odoo Enterprise Resource Planning (ERP) system storing company's transaction data. However, Odoo doesn't have a module for managing data. It takes a module for managing data into useful information.Based on the above problems, a module for data mining Market Basket Analysis is being designed. This module uses FP-Growth algorithm by utilizing the sales transaction data.For the testing, this module using data from X Supermarket. The final result of this module is an association rule from data m...

Synthesis of a basket-shaped C56H38 hydrocarbon as a precursor toward an end-cap template for carbon [6,6]nanotubes.

Science.gov (United States)

Cui, Hu; Akhmedov, Novruz G; Petersen, Jeffrey L; Wang, Kung K

2010-03-19

A basket-shaped C(56)H(38) hydrocarbon (3) possessing a 30-carbon difluorenonaphthacenyl core that can be mapped onto the surface of C(78) was synthesized from 4-bromo-1-indanone. The first stage of the synthesis involved the preparation of tetraketone 10 as a key intermediate. The use of cascade cyclization reactions of benzannulated enyne-allenes as key features in the next stage of the synthetic sequence provides an efficient route to 3 from 4-bromo-1-indanone in 12 steps. The all-cis relationship among the methyl groups and the methine hydrogens causes the two benzofluorenyl units in 3 to be in an essentially perpendicular orientation to each other. Hydrocarbon 3 and its derivatives could serve as attractive precursors leading to a geodesic C(68)H(26) end-cap template for carbon [6,6]nanotubes.

Mass transfer in fuel cells. [electron microscopy of components, thermal decomposition of Teflon, water transport, and surface tension of KOH solutions

Science.gov (United States)

Walker, R. D., Jr.

1973-01-01

Results of experiments on electron microscopy of fuel cell components, thermal decomposition of Teflon by thermogravimetry, surface area and pore size distribution measurements, water transport in fuel cells, and surface tension of KOH solutions are described.

Impregnation/Agglomeration Laboratory Tests of Heavy Fuel from Prestige to Improve Its Manageability and Removal from Seawater Surface. (Physical Behaviour of Fuel Agglomates)

International Nuclear Information System (INIS)

Garcia Frutos, F. J.; Rodriguez, V.; Otero, J.

2002-01-01

The handling and removal problems showed by heavy fuel floating in seawater could be improved or solved by using materials that agglomerate it. These materials must fulfill the following conditions: be inert materials in marine environment, the agglomerated fuel/material should float and its application and removal should be done using simple technologies. Based on these requirements, clay minerals, pine chips, mineral coal and charcoal were selected. The preliminary/results on impregnation/agglomeration with the materials mentioned above of heavy fuel from Prestige at lab scale are presented in this paper. The results have shown that only hydrophobic materials, such as mineral coal and charcoal, are able to agglomerate with fuel, which is also a hydrophobic substance. Whereas the agglomerates fuel/mineral coal sink, the agglomerates fuel/charcoal keep floating on water surface. It can be concluded that the addition of charcoal on dispersed fuel in seawater could improve its handling and removal. In this sense, pilot scale and eventually controlled in situ tests to study the feasibility of the proposed solution should be performed. (Author) 2 refs

Preliminary assessment of the benefits of derating a cask for increasing age/burnup capability

International Nuclear Information System (INIS)

Broadhead, B.L.; Parks, C.V.; Joy, D.S.; Tang, J.S.

1992-01-01

This paper discusses a study performed to determine the extent to which the age/burnup capability of the Babcock and Wilcox BR-100 rail cask could be extended by reducing the number of fuel assemblies. Only the shielding effects of derating are accounted for in this study. Separate analyses will be necessary to address the enhanced heat loads due to increased burnup or decreased age. The criterion used to assess the derating was the calculated dose 2 m from the rail car. The reference calculations were based on the 70% design of the BR-100 cask with 21 PWR fuel assemblies. Seven different basket/assembly loading configurations were investigated. The results indicate that both an alternative 18-assembly basket configuration and a 17-assembly/4-empty-hole configuration for the 21-element basket offer substantial gains over the fully loaded reference 21-element basket configuration

Fuel cladding tube and fuel rod for BWR type reactor

International Nuclear Information System (INIS)

Urata, Megumu; Mitani, Shinji.

1995-01-01

A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom (depth of the groove) in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube (depth of the groove) is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. (I.N.)

Near surface spent fuel storage: environmental issues

International Nuclear Information System (INIS)

Nelson, I.C.; Shipler, D.B.; McKee, R.W.; Glenn, R.D.

1979-01-01

Interim storage of spent fuel appears inevitable because of the lack of reprocessing plants and spent fuel repositories. This paper examines the environmental issues potentially associated with management of spent fuel before disposal or reprocessing in a reference scenario. The radiological impacts of spent fuel storage are limited to low-level releases of noble gases and iodine. Water needed for water basin storage of spent fuel and transportation accidents are considered; the need to minimize the distance travelled is pointed out. Resource commitments for construction of the storage facilities are analyzed

Experience and prospects of WWER-1000 reactor spent fuel transport

International Nuclear Information System (INIS)

Kondratyev, A.N.; Yershov, V.N.; Kozlov, Yu.V.; Kosarev, Yu.A.; Ilyin, Yu.V.; Pavlov, M.S.

1989-01-01

The paper deals with the USSR experience in shipping the commercial WWER-1000 reactor spent fuel in TK-10 and TK-13 casks. The cask designs, their basic characteristics and the WWER-1000 spent fuel features are described. An example of calculational/experimental approach in the design of a basket (one of the most important components) for spent fuel assembly (SFA) accommodation in a cask is given. The main problems of future development works are presented in brief. A concept of development of nuclear power industry with the closed fuel cycle is assumed in the Soviet Union, hence the spent nuclear fuel is to be transported from NPPs to reprocessing plants. To transport WWER-1000 spent fuel, the casks of two types were developed. These are: a pilot TK-10 cask of 3t capacity in fuel; a commercial TK-13 cask of ∼6t capacity in fuel. The pilot TK-10 cask is thick-walled (360mm) cylindrical vessel manufactured of steel shells and a bottom welded to each other. The material of the body is carbon steel. There is a steel jacket on the outer side of the cask body and at 120 mm distance off the bottom. On its cylindrical part between the jacket and the body there are T-shaped circular ribs acting as shock-absorbers. The space between the jacket and the body is filled with ethylene glycol solution of 65 degree C crystallization temperature, which functions as a neutron shielding. The TK-10 cask coolant is water or air (nitrogen) at minor excess pressure resulted from FA heatup after the cask sealing

Electrometallurgical treatment of metallic spent nuclear fuel stored at the Hanford Site

International Nuclear Information System (INIS)

Laidler, J.J.; Gay, E.C.

1996-01-01

The major component of the DOE spent nuclear fuel inventory is the metallic fuel stored at the Hanford site in the southeastern part of the state of Washington. Most of this fuel was discharged from the N-Reactor; a small part of the inventory is fuel from the early Hanford production reactors. The U.S. Department of Energy (DOE) plans to remove these fuels from the spent fuel storage pools in which they are presently stored, dry them, and place them in interim storage at a location at the Hanford site that is far removed from the Columbia River. It is not yet certain that these fuels will be acceptable for disposal in a mined geologic repository without further treatment, due to their potential pyrophoric character. A practical method for treatment of the Hanford metallic spent fuel, based on an electrorefining process, has been developed and has been demonstrated with unirradiated N-Reactor fuel and with simulated single-pass reactor (SPR) spent fuel. The process can be operated with any desired throughput rates; being a batch process, it is simply a matter of setting the size of the electrorefiner modules and the number of such modules. A single module, prototypic of a production-scale module, has been fabricated and testing is in progress at a throughput rate of 150 kg (heavy metal) per day. The envisioned production version would incorporate additional anode baskets and cathode tubes and provide a throughput rate of 333 kgHM/day. A system with four of these modules would permit treatment of Hanford metallic fuels at a rate of at least 250 metric tons per year

Analysis of heterogeneous oxygen exchange and fuel oxidation on the catalytic surface of perovskite membranes

KAUST Repository

Hong, Jongsup

2013-10-01

The catalytic kinetics of oxygen surface exchange and fuel oxidation for a perovskite membrane is investigated in terms of the thermodynamic state in the immediate vicinity of or on the membrane surface. Perovskite membranes have been shown to exhibit both oxygen perm-selectivity and catalytic activity for hydrocarbon conversion. A fundamental description of their catalytic surface reactions is needed. In this study, we infer the kinetic parameters for heterogeneous oxygen surface exchange and catalytic fuel conversion reactions, based on permeation rate measurements and a spatially resolved physical model that incorporates detailed chemical kinetics and transport in the gas-phase. The conservation equations for surface and bulk species are coupled with those of the gas-phase species through the species production rates from surface reactions. It is shown that oxygen surface exchange is limited by dissociative/associative adsorption/desorption of oxygen molecules onto/from the membrane surface. On the sweep side, while the catalytic conversion of methane to methyl radical governs the overall surface reactions at high temperature, carbon monoxide oxidation on the membrane surface is dominant at low temperature. Given the sweep side conditions considered in ITM reactor experiments, gas-phase reactions also play an important role, indicating the significance of investigating both homogeneous and heterogeneous chemistry and their coupling when examining the results. We show that the local thermodynamic state at the membrane surface should be considered when constructing and examining models of oxygen permeation and heterogeneous chemistry. © 2013 Elsevier B.V.

Analysis of heterogeneous oxygen exchange and fuel oxidation on the catalytic surface of perovskite membranes

KAUST Repository

Hong, Jongsup; Kirchen, Patrick; Ghoniem, Ahmed F.

2013-01-01

The catalytic kinetics of oxygen surface exchange and fuel oxidation for a perovskite membrane is investigated in terms of the thermodynamic state in the immediate vicinity of or on the membrane surface. Perovskite membranes have been shown to exhibit both oxygen perm-selectivity and catalytic activity for hydrocarbon conversion. A fundamental description of their catalytic surface reactions is needed. In this study, we infer the kinetic parameters for heterogeneous oxygen surface exchange and catalytic fuel conversion reactions, based on permeation rate measurements and a spatially resolved physical model that incorporates detailed chemical kinetics and transport in the gas-phase. The conservation equations for surface and bulk species are coupled with those of the gas-phase species through the species production rates from surface reactions. It is shown that oxygen surface exchange is limited by dissociative/associative adsorption/desorption of oxygen molecules onto/from the membrane surface. On the sweep side, while the catalytic conversion of methane to methyl radical governs the overall surface reactions at high temperature, carbon monoxide oxidation on the membrane surface is dominant at low temperature. Given the sweep side conditions considered in ITM reactor experiments, gas-phase reactions also play an important role, indicating the significance of investigating both homogeneous and heterogeneous chemistry and their coupling when examining the results. We show that the local thermodynamic state at the membrane surface should be considered when constructing and examining models of oxygen permeation and heterogeneous chemistry. © 2013 Elsevier B.V.

Linear programming to build food-based dietary guidelines: Romanian food baskets

DEFF Research Database (Denmark)

Parlesak, Alexandr; Robertson, Aileen; Hondru, Gabriela

approach using linear programming methodology to design national dietary recommendations which aim to prevent both NCDs and micronutrient deficiencies and still be affordable by low income groups. This new approach is applied within the context of food availability in Romania in 2014. Eating the same food...... every day is unrealistic and too monotonous to be maintained, so this novel approach is used to select a wide range of diverse foods that can be recommended for a period of up to, for example, one month. The following are the key findings of this report. • The simplest version of the Romanian food.......65 lei (~€ 4.46) for a day. • Key nutrients, primarily vitamin D, calcium, potassium and iron, were found to control the overall price. • The least expensive basket (one day’s rations) is monotonous and the linear programming approach is used to select a wide range of foods that can be recommended...

Conceptual Assessment of a Fresh Fuel Transport Package for KJRR

Energy Technology Data Exchange (ETDEWEB)

Lee, Ju-Chan; Choi, W. S.; Bang, K. S.; Yu, S. H.; Park, J. S.; Yang, Y. Y. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The IAEA and domestic regulations stipulate that the fissile material transport package be subjected to the cumulative effects of a 9 m drop, 1 m puncture, 800 ℃ thermal and water leakage tests. A fissile material transport package should be maintained the subcriticality during the normal and accident conditions for contingency of leakage of water into or out of package, rearrangement of the contents, reduction of spaces and temperature changes. KAERI has been developing a fresh fuel transport package for Kijang research reactor (KJRR). This paper describes a conceptual design and preliminary safety analysis of the transport package for KJRR. The transport package was designed for shipment of a fresh fuel and a FM (Fission Molybdenum) target. Low-enriched uranium (LEU) of U-Mo fuel with U-235 enrichment of 19.75 w/o is used as a research reactor fuel. And LEU of UAlx-Al with U-235 enrichment of 19.75 w/o is used as a FM target material. The transport package was designed for shipment of a fresh fuel and a FM target. Safety analyses were conducted on all areas, including criticality, structural, and thermal fields. In the criticality analysis, effective neutron multiplication factors were below the criticality safety limit. In the structural analysis, the maximum stress satisfied the stress requirement stipulated in the ASME code. After 9 m free drop and 1 m puncture test, there was no significant deformation of fuel basket to cause a criticality. In the thermal analysis, the maximum temperatures at each part were lower than the allowable values.

Numerically predicting horizontally oriented spent fuel rod surface temperatures

International Nuclear Information System (INIS)

Wix, S.D.; Koski, J.A.

1993-01-01

A comparison between numerical calculations with use of commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degrees C and 23 degrees C for the low heat dissipation and high dissipation, respectively. The temperature predictions using helium as a fill gas are lower than the experimental data for the low and medium heat dissipation levels. The temperature predictions are 1 degrees C and 6 degrees C lower than the experimental data for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degrees C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include a experimental uncertainity in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This works demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects, such as axial heat transfer through the spent fuel rods, will be increasingly important as the amount of dissipated heat increases

Numerically predicting horizontally oriented spent fuel rod surface temperatures

International Nuclear Information System (INIS)

Wix, S.D.; Koski, J.A.

1992-01-01

A comparison between numerical calculations with use of commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degree C and 23 degree C for the low heat dissipation and high heat dissipation, respectively. The temperature predictions using helium as a fill gas are lower than the experimental data for the low and medium heat dissipation levels. The temperature predictions are 1 degree C and 6 degree C lower than the experimental data for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degree C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include experimental uncertainty in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This work demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects, such as axial heat transfer through the spent fuel rods, will be increasingly important as the amount of dissipated heat increases

Influencia del tamaño del balón y la altura de la canasta en la eficacia del lanzamiento en categoría pre-infantil (11-12 años femenina.[Influence of ball size and basket height in the effectiveness of the shooting in femalepre-child level (11-12 years

Directory of Open Access Journals (Sweden)

María Teresa Gutiérrez Colón

2010-03-01

Abstract The aim of this study is to know which are learning best conditions of the shooting basketball in the phase pre-child. In this phase is a sharp change for the boy depending on the size of the equipment, that demands to carry out a progressive adaptation process of dimensions of the ball and of the basket, allowing that players have a more suitable and adapted learning to his characteristics. The sample was formed by 12 players pre-child (11-12 years, belonging an equipment from feminine basketball. The effectiveness of the releases was analyzed to basket depending on the size of the ball and the height of the basket, as well as of the position of the shooting, before and after an intervention process. Results showed that at the start of the process were differences in the effectiveness of shooting. Players were more effective shooting with ball of minibasket (number 5 and basket of minibasket that with those of the full-grown game (number 6. Finally, after the intervention process results showed as players increase their effectiveness when they shoot with big ball and basket with respect to the beginning of the process. Besides, in the post-test obtained great effectiveness values with this material that with the small ball and basket. Key words:  basketball, pre-child, effectiveness, shooting, ball size, basket

The influence of near field hydrogen on actinide solubilities and spent fuel leaching

International Nuclear Information System (INIS)

Spahiu, K.; Werme, L.; Eklund, U.B.

2000-01-01

Large amounts of hydrogen are produced as a result of the anoxic corrosion of iron in the proposed container materials for some geologic repositories. Another hydrogen source, less important than the anoxic corrosion of iron, is the radiolysis of water by the spent fuel radiation. Gas phase formation occurs when the pressure of the hydrogen equals at least the hydrostatic pressure, around 5 MPa at 500 meters depth. The effects of 5 MPa hydrogen pressure on spent PWR fuel leaching and on uranium oxide solubility have been studied in carbonated solutions at 70 C. The experiments were performed in a 1 liter autoclave, filled with 950 ml of a solution 10 mM NaCl, 2 mM NaHCO 3 and with hydrogen at a pressure of 5 MPa in the remaining 50 ml free volume. The leaching behavior of 2 g PWR spent fuel powder of the 0.25-0.50 mm fraction, placed in a gold basket was studied during several months by analyzing 10 ml solution samples taken after regular time intervals. A few experiments were performed also with unirradiated U(IV) oxide. In both cases extremely low concentrations of uranium (less than 10 -9 M) were measured in the solution samples. Furthermore the uranium levels in solution remained practically constant during the whole leaching period (more than one year), indicating the absence of any oxidative dissolution of the spent fuel matrix. The same conclusion is confirmed by the constant (within analytical errors) levels of strontium, cesium, molybdenum, iodine and technetium during the whole leaching period. These results have been compared with the ones obtained during the leaching of a spent fuel pin in anoxic conditions, where the uranium and other radionuclides levels are several orders of magnitude higher. The surface of spent fuel or U(IV) oxide is partially oxidized during storage, giving rise to relatively high levels of U(VI) in solution even during leaching in anoxic conditions. No such effect could be observed in the presence of 5 MPa hydrogen, indicating

Modeling fuel succession

Science.gov (United States)

Brett Davis; Jan van Wagtendonk; Jen Beck; Kent van Wagtendonk

2009-01-01

Surface fuels data are of critical importance for supporting fire incident management, risk assessment, and fuel management planning, but the development of surface fuels data can be expensive and time consuming. The data development process is extensive, generally beginning with acquisition of remotely sensed spatial data such as aerial photography or satellite...

National Beef Market Basket Survey - 2006: External fat thickness measurements and separable component determinations for beef from US retail establishments.

Science.gov (United States)

Mason, C L; Nicholson, K L; Brooks, J C; Delmore, R J; Henning, W R; Johnson, D D; Lorenzen, C L; Maddock, R J; Miller, R K; Morgan, J B; Wasser, B E; Gwartney, B L; Harris, K B; Griffin, D B; Hale, D S; Savell, J W

2009-02-01

A market basket survey for beef retail cut composition at the retail level (four stores each from two chains in each city) was conducted in 11 US cities from January to March 2006. Beef cuts (n=17,495) were measured for external fat thickness with cuts from the chuck (0.05cm), round (0.05cm), and miscellaneous (0.04cm) having less (Pmarketing purposes.

Apparatus for loading fuel pellets in fuel rods

International Nuclear Information System (INIS)

Tedesco, R.J.

1976-01-01

An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

Transport of Spent Nuclear Fuels, High and Intermediate Level Wastes: A Continuous Challenge

International Nuclear Information System (INIS)

Otton, C.; Blachet, L.

2009-01-01

For more than 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the used nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. In this presentation we will focus on the casks for the spent fuel, high level waste and intermediate level waste transportation. Answering to the constant evolution of the nuclear industry transport needs is a challenge that TN International faces routinely. Concerning the spent nuclear fuel transportation, TN International has developed in the early 80's a fleet of TN12 type casks fitted with several types of baskets able to safely transport all the spent fuel from the nuclear power plant or the research laboratories to AREVA La Hague plant. The current challenge is the design of a new transport cask generation taking into account the needs of the industry for the next 30 years. The replacement of the TN12 cask generation is to be scheduled as the regulations have changed and the fuel characteristics have evolved. The new generation of casks will take into account all the technical evolutions made during the TN12 thirty years of use. MOX spent fuel has now its dedicated cask: the TN112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in 2008 in the EDF nuclear power plant of Saint-Laurent-des-Eaux. Concerning the high level waste such as the La Hague vitrified residues a whole fleet of casks has been developed such as the TN 28 VT dedicated to transport, the TN81 and TN85 dedicated to transport and storage. These casks have permitted the

A methodology for the evaluation of fuel rod failures under transportation accidents

International Nuclear Information System (INIS)

Rashid, J.Y.R.; Machiels, A.J.

2004-01-01

Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

A basket two-part model to analyze medical expenditure on interdependent multiple sectors.

Science.gov (United States)

Sugawara, Shinya; Wu, Tianyi; Yamanishi, Kenji

2018-05-01

This study proposes a novel statistical methodology to analyze expenditure on multiple medical sectors using consumer data. Conventionally, medical expenditure has been analyzed by two-part models, which separately consider purchase decision and amount of expenditure. We extend the traditional two-part models by adding the step of basket analysis for dimension reduction. This new step enables us to analyze complicated interdependence between multiple sectors without an identification problem. As an empirical application for the proposed method, we analyze data of 13 medical sectors from the Medical Expenditure Panel Survey. In comparison with the results of previous studies that analyzed the multiple sector independently, our method provides more detailed implications of the impacts of individual socioeconomic status on the composition of joint purchases from multiple medical sectors; our method has a better prediction performance.

A food basket investigation during the autumn of 1994

International Nuclear Information System (INIS)

Moere, H.; Falk, R.; Svedjemark, G.A.; Becker, W.; Brugaard Konde, Aa.

1995-10-01

During the autumn of 1994 an investigation of foodstuffs has been accomplished to assess the average intake of 137 Cs by the Swedish population due to the Chernobyl accident. A standardized food basket has been collected from two grocers in 10 localities, of which the majority came from areas with the highest fallout. The estimated maximum intake of 137 Cs was 815 Bq/year in the inland of the county of Vaesterbotten. The population weighted average intake for the fallout affected counties was 435 Bq/year. The rest of the county received an intake of 235 Bq/year. The population weighted average of the intake for the whole county was estimated to 274 Bq/year. From this intake the calculated body burden would be 1.3 Bq/kg for the average citizen. Whole-body measurements of a sample of the population gave 2.0 Bq/kg. A plausible explanation would be that 40% of the intake of 137 Cs can have its origin from the 10% of the consumption of foodstuffs that are home produced or collected for the average individual in Sweden. The average intake of 274 Bq/year gives a committed effective dose equivalent of 3.6 μSv. 6 refs, 10 tabs

Development of surface enhanced Raman scattering (SERS) spectroscopy monitoring of fuel markers to prevent fraud

Science.gov (United States)

Wilkinson, Timothy; Clarkson, John; White, Peter C.; Meakin, Nicholas; McDonald, Ken

2013-05-01

Governments often tax fuel products to generate revenues to support and stimulate their economies. They also subsidize the cost of essential fuel products. Fuel taxation and subsidization practices are both subject to fraud. Oil marketing companies also suffer from fuel fraud with loss of legitimate sales and additional quality and liability issues. The use of an advanced marking system to identify and control fraud has been shown to be effective in controlling illegal activity. DeCipher has developed surface enhanced Raman scattering (SERS) spectroscopy as its lead technology for measuring markers in fuel to identify and control malpractice. SERS has many advantages that make it highly suitable for this purpose. The SERS instruments are portable and can be used to monitor fuel at any point in the supply chain. SERS shows high specificity for the marker, with no false positives. Multiple markers can also be detected in a single SERS analysis allowing, for example, specific regional monitoring of fuel. The SERS analysis from fuel is also quick, clear and decisive, with a measurement time of less than 5 minutes. We will present results highlighting our development of the use of a highly stable silver colloid as a SERS substrate to measure the markers at ppb levels. Preliminary results from the use of a solid state SERS substrate to measure fuel markers will also be presented.

Concrete spent fuel storage casks dose rates

International Nuclear Information System (INIS)

Bace, M.; Jecmenica, R.; Trontl, K.

1998-01-01

Our intention was to model a series of concrete storage casks based on TranStor system storage cask VSC-24, and calculate the dose rates at the surface of the casks as a function of extended burnup and a prolonged cooling time. All of the modeled casks have been filled with the original multi-assembly sealed basket. The thickness of the concrete shield has been varied. A series of dose rate calculations for different burnup and cooling time values have been performed. The results of the calculations show rather conservative original design of the VSC-24 system, considering only the dose rate values, and appropriate design considering heat rejection.(author)

Atomic layer deposition of ruthenium surface-coating on porous platinum catalysts for high-performance direct ethanol solid oxide fuel cells

Science.gov (United States)

Jeong, Heon Jae; Kim, Jun Woo; Jang, Dong Young; Shim, Joon Hyung

2015-09-01

Pt-Ru bi-metallic catalysts are synthesized by atomic layer deposition (ALD) of Ru surface-coating on sputtered Pt mesh. The catalysts are evaluated in direct ethanol solid oxide fuel cells (DESOFCs) in the temperature range of 300-500 °C. Island-growth of the ALD Ru coating is confirmed by transmission electron microscopy and X-ray photoelectron spectroscopy (XPS) analyses. The performance of the DESOFCs is evaluated based on the current-voltage output and electrochemical impedance spectroscopy. Genuine reduction of the polarization impedance, and enhanced power output with improved surface kinetics are achieved with the optimized ALD Ru surface-coating compared to bare Pt. The chemical composition of the Pt/ALD Ru electrode surface after fuel cell operation is analyzed via XPS. Enhanced cell performance is clearly achieved, attributed to the effective Pt/ALD Ru bi-metallic catalysis, including oxidation of Cdbnd O by Ru, and de-protonation of ethanol and cleavage of C-C bonds by Pt, as supported by surface morphology analysis which confirms formation of a large amount of carbon on bare Pt after the ethanol-fuel-cell test.

Ignition probability of fine dead surface fuels of native Patagonian forests or Argentina

Directory of Open Access Journals (Sweden)

Lucas O. Bianchi

2014-04-01

Full Text Available Aim of study: The Canadian Forest Fire Weather Index (FWI is being implemented all over the world. This index is being adapted to the Argentinean ecosystems since the year 2000. With the objective of calibrating the Fine Fuel Moisture Code (FFMC of the FWI system to Patagonian forests, we studied the relationship between ignition probability and fine dead surface fuel moisture content (MC as an indicator of potential fire ignition.Area of study: The study area is located in northwestern Patagonia, Argentina, and comprised two main forest types (cypress and ñire grown under a Mediterranean climate, with a dry summer and precipitations during winter and autumn (~500-800 mm per year.Material and Methods: We conducted lab ignition tests fires to determine the threshold of fine dead fuel ignition at different MC levels. Moisture content of dead fine surface fuels in the field was measured every 10-15 days from November to March for three seasons. We calculated the FFMC during these seasons and correlated it with the measured MC by applying a logistic regression model. We combined the results of the ignition tests and of the regressions to suggest FFMC categories for estimating fire danger in Patagonian forests.Main results: The ignition threshold occurred at MC values of 21.5 and 25.0% for cypress and ñire sites, respectively. The MC measured varied from 7.3 to 129.6%, and the calculated FFMC varied between 13.4 and 92.6. Highly significant regressions resulted when FFMC was related to MC. The ignition threshold corresponded to a FFMC=85. We proposed to divide the FFMC scale in three fire danger categories: Low (FFMC≤85, High (8589.Research highlights: Our results provide a useful tool for predicting fire danger in these ecosystems, and are a contribution to the development of the Argentinean Fire Danger Rating and a reference for similar studies in other countries where the FWI is being implemented

The coupling effect of gas-phase chemistry and surface reactions on oxygen permeation and fuel conversion in ITM reactors

KAUST Repository

Hong, Jongsup

2015-08-01

© 2015 Elsevier B.V. The effect of the coupling between heterogeneous catalytic reactions supported by an ion transport membrane (ITM) and gas-phase chemistry on fuel conversion and oxygen permeation in ITM reactors is examined. In ITM reactors, thermochemical reactions take place in the gas-phase and on the membrane surface, both of which interact with oxygen permeation. However, this coupling between gas-phase and surface chemistry has not been examined in detail. In this study, a parametric analysis using numerical simulations is conducted to investigate this coupling and its impact on fuel conversion and oxygen permeation rates. A thermochemical model that incorporates heterogeneous chemistry on the membrane surface and detailed chemical kinetics in the gas-phase is used. Results show that fuel conversion and oxygen permeation are strongly influenced by the simultaneous action of both chemistries. It is shown that the coupling somewhat suppresses the gas-phase kinetics and reduces fuel conversion, both attributed to extensive thermal energy transfer towards the membrane which conducts it to the air side and radiates to the reactor walls. The reaction pathway and products, in the form of syngas and C2 hydrocarbons, are also affected. In addition, the operating regimes of ITM reactors in which heterogeneous- or/and homogeneous-phase reactions predominantly contribute to fuel conversion and oxygen permeation are elucidated.

Thermalhydraulic analyses of AECL's spent fuel dry storage systems

International Nuclear Information System (INIS)

Moffett, R.; Sabourin, G.

1995-01-01

This paper presents the validation of one- and three-dimensional thermalhydraulic models to be used to evaluate the thermal performance of AECL's MACSTOR and CANSTOR spent fuel dry storage modules. For this purpose, we compared analytical results to results of experiments conducted at AECL's Whiteshell Laboratories where mockups of the MACSTOR module and of a CANDU fuel storage basket were tested. The paper shows improvements to a simple one-dimensional model of the MACSTOR mock-up used previously. The replacement of constant heat transfer coefficients by free convection correlations, the addition of a storage cylinder model, and the addition of a radiation heat transfer model improved the predictions of concrete and storage cylinder temperatures. The paper also presents a new three-dimensional model for flow and heat transfer in the MACSTOR mock-up developed using CFDS-FLOW3D and -RAD3D computer programs. CFDS-FLOW3D code can estimate loss coefficients in complex geometry to an accuracy better than standard engineering correlations. The flow and temperature fields predicted using CFDS-FLOW3D are consistent with the measurements made during MACSTOR mock-up experiments (author). 5 refs., 4 tabs., 9 figs

Sudden oak death-caused changes to surface fuel loading and potential fire behavior in Douglas-fir-tanoak forests

Science.gov (United States)

Y.S. Valachovic; C.A. Lee; H. Scanlon; J.M. Varner; R. Glebocki; B.D. Graham; D.M. Rizzo

2011-01-01

We compared stand structure and fuel loading in northwestern California forests invaded by Phytophthora ramorum, the cause of sudden oak death, to assess whether the continued presence of this pathogen alters surface fuel loading and potential fire behavior in ways that may encumber future firefighting response. To attempt to account for these...

Physical and chemical characteristics of surface fuels in masticated mixed-conifer stands of the U.S. Rocky Mountains

Science.gov (United States)

Robert E. Keane; Pamela G. Sikkink; Theresa B. Jain

2018-01-01

Mastication is a wildland fuel treatment technique that is rapidly becoming the preferred method for many fire hazard reduction projects, especially in areas where reducing fuels with prescribed fire is particularly challenging. Mastication is the process of mechanically modifying the live and dead surface and canopy biomass by chopping and shredding vegetation to...

The technique for determination of surface contamination by uranium on U3Si2-Al plate-type fuel elements

International Nuclear Information System (INIS)

Li Shulan; He Fengqi; Wang Qingheng; Han Jingquan

1993-04-01

The NDT method for determining the surface contamination by uranium on U 3 Si 2 -Al plate-type fuel elements, the process of standard specimen preparation and the graduation curve are described. The measurement results of U 3 Si 2 -Al plate-type fuel elements show that the alpha counting method to measure the surface contamination by uranium on fuel plate is more reliable. The UB-1 type surface contamination meter, which was recently developed, has many advantages such as high sensitivity to determine the uranium pollution, short time in measuring, convenience for operation, and the minimum detectable amount of uranium is 5 x 10 -10 g/cm 2 . The measuring device is controlled by a microcomputer. Besides data acquisition and processing, it has functions of statistics, output data on terminal or to printer and alarm. The procedures of measurement are fully automatic. All of these will meet the measuring needs in batch process

Design of Product Placement Layout In Retail Shop Using Market Basket Analysis

Directory of Open Access Journals (Sweden)

Isti Surjandari

2010-10-01

Full Text Available Retailing is an industry with high level of competition.  It is a customer-based industry which depends on how it could be aware of what the customers' needs and requirements are. One technique most used in supermarkets is the mix merchandise. The purpose of this paper is to identify associated products, which then grouped in mix merchandise with the use of market basket analysis. This association between products then will be applied in the design layout of the product in the supermarket. The process of identifying the related products bought together in one transaction is done by using data mining  technique. Apriori algorithm is chosen as a method in the data mining process. Using WEKA (Waikato Environment for Knowledge Analysis software, the association rule between products is calculated. The results found five category association rules and fourteen sub-category association rules. These associations then will be interpreted as confidence and support to become consideration for the product layout.

Inspecting method for fuel rods

International Nuclear Information System (INIS)

Watanabe, Masaaki; Kogure, Sumio.

1976-01-01

Purpose: To precisely detect the response of flaw in clad tube and submerged fuel pellets from a relationship between the surface of fuel rod and internal signal. Constitution: Ultrasonic reflected waves from the surface of fuel rods and the interior are detected and either one of fuel rod or ultrasonic flaw detecting contact is rotated to thereby precisely detect the response of the flaw of clad tube and submerged fuel pellets from a relationship between said surface and the interior. It will be noted that the ultrasonic flaw detecting contact used is of the line-focus type, the incident angle of ultrasonic wave from the ultrasonic flaw detecting contact relative to the fuel rod is the angle of skew, that is, the ultrasonic flaw detecting contact is not perpendicular to a center axis of the fuel rod but is slightly displace. That is, the use of the aforesaid contact may facilitate discrimination between the surface flaw of the fuel rod and the response of submergence, and in addition, the employment of the aforesaid incident angle makes it hard to receive reflected waves from the surface of the fuel rod which is great in terms of energy to facilitate discrimination of waves responsive to submergence. (Kawakami, Y.)

Hazards classification determination for PUREX fuel transfer to K-Basins

International Nuclear Information System (INIS)

Dodd, E.N. III.

1995-01-01

The PUREX Plant presently contains 2.9 metric tons of an aluminum clad Single Pass Reactor (SPR) fuel which is stored under water in four open top buckets in the PUREX slug storage basin. The PUREX dissolver cells contain approximately 0.5 metric tons of zirconium clad N Reactor fuel which was inadvertently placed into the process cell during charging operations. The dissolver N reactor elements will be recovered from the process floors using new crane operated tools. When the fuel shipment(s) is scheduled, the cask cars will be positioned into the PUREX rail tunnel and the overhead door will be opened. All the SPR fuel will be loaded into two cask rail cars inside four casks. The N Reactor fuel will be loaded into a separate rail car inside two or three casks. The car loading is initiated by opening the rail car lid and removing the cask lids. Prior to loading the canisters of N Reactor fuel, the canisters will be refilled with water (as needed) and a lid will be installed. The baskets of SPR fuel or canisters of N Reactor fuel will then be loaded into the casks. The lids to the casks will then be reinstalled and the car lids closed. The rail cars will then be decontaminated as necessary. The cask cars will be shipped either in two shipments or a combined single shipment using the rail route between PUREX and the K Basins. At the basin, the cask car will be positioned in the loadout area. The cask car lid will be opened and a single cask moved into the loadout pit, which is a lowered section of the basin. The cask lid is removed while the cask is lower into the pit. The fuel is then removed from the cask and stored in the basin. The cask is then removed, the lid reinstalled during removal, and the cask replaced into the cask car. This document identifies the hazard classification of the Fuel Transfer from the PUREX facility to K-Basins

Containing method for spent fuel and spent fuel containing vessel

International Nuclear Information System (INIS)

Maekawa, Hiromichi; Hanada, Yoshine.

1996-01-01

Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

(02.2) Scoping experiments; (02.3) long-term corrosion testing and properties evaluation of candidate waste package basket material

International Nuclear Information System (INIS)

VanKonynenburg, R. A.

1996-01-01

The work described in this activity plan addresses Information Need 2.7.3 of the Yucca Mountain Site Characterization Plan (l), which reads Determination that the design criteria in lOCFR60.130 through 60.133 and any appropriate additional design objectives pertaining to criticality control have been met. This work falls under section WBS 1.2.2.5 2 (Basket Materials) of WBS 1.2.2.5 (Waste Package Materials) in the Work Breakdown Structure of the Yucca Mountain Site Characterization Project

Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

International Nuclear Information System (INIS)

1995-07-01

The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D ampersand D) and to reduce the cost of maintaining the facilities prior to D ampersand D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor's fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered

Fuel-to-cladding heat transfer coefficient into reactor fuel element

International Nuclear Information System (INIS)

Lassmann, K.

1979-01-01

Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

Fuel charging machine

International Nuclear Information System (INIS)

Uchikawa, Sadao.

1978-01-01

Purpose: To enable continuous fuel discharging and charging steps in a bwr type reactor by effecting positioning only for once by providing a plurality of fuel assembly grippers and their drives co-axially on a rotatable surface. Constitution: A plurality of fuel assembly grippers and their drives are provided co-axially on a rotatable surface. For example, a gripper A, a drive B, a gripper C and a drive D are arranged co-axially in symmetric positions on a disk rotated on rails by wheels and rotational drives. A new fuel in a fuel pool is gripped by the gripper A and transported above the reactor core. Then, the disk is positioned so that the gripper C can grip the spent fuel in the core, and the fuel to be discharged is gripped and raised by the gripper C. Then the disk is rotated by 180 0 and the new fuel in the gripper A is charged into the position from which the old fuel has been discharged and, finally, the discharged fuel is sent to the fuel pool for storage. (Seki, T.)

Nuclear fuel assembly

International Nuclear Information System (INIS)

Takeda, Tadashi; Sato, Kenji; Goto, Masakazu.

1984-01-01

Purpose: To facilitate identification of a fuel assembly upon fuel exchange in BWR type reactors. Constitution: Fluorescent material is coated or metal plating is applied to the impressed portion of a upper tie plate handle of a fuel assembly, and the fluorescent material or the metal plating surface is covered with a protective membrane made of transparent material. This enables to distinguish the impressed surface from a distant place and chemical reaction between the impressed surface and the reactor water can be prevented. Furthermore, since the protective membrane is formed such that it protrudes toward the upper side relative to the impressed surface, there is no risk of depositions of claddings thereover. (Moriyama, K.)

Fuel R and D to improve fuel reliability

International Nuclear Information System (INIS)

Yang, Rosa; Cheng, Bo; Deshon, Jeff; Edsinger, Kurt; Ozer, Odelli

2006-01-01

Light water reactor fuel is operating in an increasingly challenging environment. Fuel burnup extension and cycle length increase both can increase the local duty. Reactor water chemistry modifications for the purpose of protection the plant system materials have the potential of increasing fuel surface deposition and cladding corrosion and hydriding. The status of fuel performance in US reactors is summarized and an update of the Fuel Reliability Program' established by the utility industry to ensure reliability is provided. (author)

Ultra low injection angle fuel holes in a combustor fuel nozzle

Science.gov (United States)

York, William David

2012-10-23

A fuel nozzle for a combustor includes a mixing passage through which fluid is directed toward a combustion area and a plurality of swirler vanes disposed in the mixing passage. Each swirler vane of the plurality of swirler vanes includes at least one fuel hole through which fuel enters the mixing passage in an injection direction substantially parallel to an outer surface of the plurality of swirler vanes thereby decreasing a flameholding tendency of the fuel nozzle. A method of operating a fuel nozzle for a combustor includes flowing a fluid through a mixing passage past a plurality of swirler vanes and injecting a fuel into the mixing passage in an injection direction substantially parallel to an outer surface of the plurality of swirler vanes.

Examination of the surface coatings removed from K-East Basin fuel elements

International Nuclear Information System (INIS)

Abrefah, J.; Marschman, S.C.; Jenson, E.D.

1998-05-01

This report provides the results of studies conducted on coatings discovered on the surfaces of some N-Reactor spent nuclear fuel (SNF) elements stored at the Hanford K-East Basin. These elements had been removed from the canisters and visually examined in-basin during FY 1996 as part of a series of characterization tests. The characterization tests are being performed to support the Integrated Process Strategy developed to package, dry, transport, and store the SNF in an interim storage facility on the Hanford site. Samples of coating materials were removed from K-East canister elements 2350E and 2540E, which had been sent, along with nine other elements, to the Postirradiation Testing Laboratory (327 Building) for further characterization following the in-basin examinations. These coating samples were evaluated by Pacific Northwest National Laboratory using various analytical methods. This report is part of the overall studies to determine the drying behavior of corrosion products associated with the K-Basin fuel elements. Altogether, five samples of coating materials were analyzed. These analyses suggest that hydration of the coating materials could be an additional source of moisture in the Multi-Canister Overpacks being used to contain the fuel for storage

Examination of the surface coating removed from K-East Basin fuel elements

Energy Technology Data Exchange (ETDEWEB)

Abrefah, J.; Marschman, S.C.; Jenson, E.D.

1998-05-01

This report provides the results of studies conducted on coatings discovered on the surfaces of some N-Reactor spent nuclear fuel (SNF) elements stored at the Hanford K-East Basin. These elements had been removed from the canisters and visually examined in-basin during FY 1996 as part of a series of characterization tests. The characterization tests are being performed to support the Integrated Process Strategy developed to package, dry, transport, and store the SNF in an interim storage facility on the Hanford site. Samples of coating materials were removed from K-East canister elements 2350E and 2540E, which had been sent, along with nine other elements, to the Postirradiation Testing Laboratory (327 Building) for further characterization following the in-basin examinations. These coating samples were evaluated by Pacific Northwest National Laboratory using various analytical methods. This report is part of the overall studies to determine the drying behavior of corrosion products associated with the K-Basin fuel elements. Altogether, five samples of coating materials were analyzed. These analyses suggest that hydration of the coating materials could be an additional source of moisture in the Multi-Canister Overpacks being used to contain the fuel for storage.

The surface water model for assessing Canada's nuclear fuel waste disposal concept

International Nuclear Information System (INIS)

Bird, G.A.; Stephenson, M.; Cornett, R.J.

1993-01-01

Canada's Nuclear Fuel Waste Management Program (NFWMP) is investigating the concept of disposal of nuclear fuel waste in a vault excavated deep in crystalline rock on the Canadian Shield. Probabilistic vault, geosphere, and biosphere models are implemented using Monte Carlo simulation techniques to trace nuclides transported in groundwater to the surface environment and humans far into the future. This paper describes the surface water submodel and its parameter values, sensitivity analysis, and validation. The surface water model is a simple, time-dependent, mass balance model of a lake that calculates radioactive and stable isotope contaminant concentrations in lake water and sediment. These concentrations are input to the other submodels and used to predict the radiological dose to humans and other biota. Parameter values in the model are based on the literature and the author's own data, and are generic to Canadian Shield lakes. Most parameters are represented by log normally distributed probability density functions. Sensitivity analysis indicates that nuclide concentrations in lake water and sediment are governed primarily by hydrological flushing with catchment area being the most important parameter. When catchment area is held constant lake area and nuclide transfer rate from water to sediment strongly influence concentrations in both water and sediment. For volatile nuclides, gaseous evasion also has a marked influence on concentrations in both water and sediment, whereas sedimentation rate strongly influences sediment nuclide concentrations. Validation tests demonstrate that the models predictions for 60 Co, 134 Cs, 3 H, P, Cd and Ca are consistent with empirical data when uncertainties are taken into account

Thermalhydraulic analyses of AECL`s spent fuel dry storage systems

Energy Technology Data Exchange (ETDEWEB)

Moffett, R; Sabourin, G [Atomic Energy of Canada Ltd., Montreal, PQ (Canada). CANDU Operations; Banas, A O [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

This paper presents the validation of one- and three-dimensional thermalhydraulic models to be used to evaluate the thermal performance of AECL`s MACSTOR and CANSTOR spent fuel dry storage modules. For this purpose, we compared analytical results to results of experiments conducted at AECL`s Whiteshell Laboratories where mockups of the MACSTOR module and of a CANDU fuel storage basket were tested. The paper shows improvements to a simple one-dimensional model of the MACSTOR mock-up used previously. The replacement of constant heat transfer coefficients by free convection correlations, the addition of a storage cylinder model, and the addition of a radiation heat transfer model improved the predictions of concrete and storage cylinder temperatures. The paper also presents a new three-dimensional model for flow and heat transfer in the MACSTOR mock-up developed using CFDS-FLOW3D and -RAD3D computer programs. CFDS-FLOW3D code can estimate loss coefficients in complex geometry to an accuracy better than standard engineering correlations. The flow and temperature fields predicted using CFDS-FLOW3D are consistent with the measurements made during MACSTOR mock-up experiments (author). 5 refs., 4 tabs., 9 figs.

Multicanister overpack topical report

Energy Technology Data Exchange (ETDEWEB)

Lorenz, B.D., Fluor Daniel Hanford

1997-03-25

The Spent Nuclear Fuel MCO is a single-use container that consists of a cylindrical shell, five to six fuel baskets, a shield plug, and features necessary for maintaining the structural integrity of the MCO while providing criticality control and fuel processing capability.

Safety aspects of the RA-6 spent fuel shipment to the USA

International Nuclear Information System (INIS)

Novara, Oscar; Facchini, Guillermo; Fernandez, Carlos

2008-01-01

RA-6 reactor is located in Bariloche Atomic Centre (CAB), in the city of San Carlos de Bariloche, in the south of Argentina. In 2005, CNEA and DOE signed a contract for the conversion of the RA-6 reactor to LEU and for shipping back in a single shipment the HEU spent fuel inventory that consisted of 42 MTR - type fuel assemblies. The shipment activity was performed in the frame of the DOE's Spent Fuel Acceptance Program. The shipment campaign took place in the last quarter of 2007 and the receiving facility for the RA-6 fuel was Savannah River Site. One unit of a NAC - LWT shipping cask was used to ship the fuel. In order to place inside it all the fuel assemblies, cropping of their non active parts (structural parts) was required. In order to provide adequate shielding to the operators, fuel cropping was performed under water. Transfer of baskets loaded with conditioned fuel to the transport cask was made by shielded intermediate transfer systems. Especially designed shielded drums were manufactured for the storage of the cropped parts that remained in the reactor site as medium-level radioactive waste. After testing of the loaded LWT (radionuclide sampling test, helium test), the package check out was completed by measuring the superficial contamination (α and β/γ emitters) and the dose rate in contact and at 1 m. An additional requirement was to verify that the package was 'self-protected'. The ISO containers with the package and with the auxiliary equipment were also subjected to an equivalent radiological control. The typical daily staff that participated in the loading campaign was about twelve people. The collective dose was 0.72 mSv.man. (author)

Spent fuel storage and isolation

International Nuclear Information System (INIS)

Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

1979-02-01

The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

Alternative Concept to Enhance the Disposal Efficiency for CANDU Spent Fuel Disposal System

International Nuclear Information System (INIS)

Lee, Jong Youl; Cho, Dong Geun; Kook, Dong Hak; Lee, Min Soo; Choi, Heui Joo

2011-01-01

There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over 100 .deg. C were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

Brief draft on surface and subsurface storage of high level and long-lived radioactive wastes. Spent fuels synthesis file

International Nuclear Information System (INIS)

Dumas, C.; Jaecki, P.

2002-01-01

This document makes a synthesis of the results of two brief draft studies performed in 2002 about the surface and subsurface storage of spent fuels. These studies stress on the long duration aspect of the disposal: feasibility of a secular disposal facility, potential risks and safety level of such a facility, estimation of the initial investment and of operation and maintenance costs. The main points of the specifications and the input data are presented first, and then the subsurface and surface draft studies are described. Content: specifications (imposed design principles and options, dry corrosion, input data); subsurface storage (description and design options, thermal dimensioning and ventilation, geotechnical stability of the facility, subsurface water management, dry corrosion, infrastructure durability, safety, monitoring, security and physical protection, technical-economical aspects, case of Mox fuel, case of glass packages); surface storage (description and design options, thermal dimensioning and ventilation, mechanical dimensioning of the facility, dry corrosion, infrastructure durability, safety, monitoring, security and physical protection, technical-economical aspects, case of Mox fuel, case of glass packages); conclusions and perspectives. (J.S.)

Calculations of radiation levels during reactor operations for safeguard inspections

International Nuclear Information System (INIS)

Sobhy, M.

1996-01-01

When an internal core spent fuel storage is used in the shield tank to accommodate a large number of spent fuel baskets, physical calculations are performed to determine the number of these spent fuel elements which can be accommodated and still maintain the gamma activity outside under the permissible limit. The corresponding reactor power level is determined. The radioactivity calculations are performed for this internal storage at different axial levels to avoid the criticality of the reactor core. Transport theory is used in calculations based on collision probability for multi group cell calculations. Diffusion theory is used in three dimensions in the core calculations. The nuclear fuel history is traced and radioactive decay is calculated, since reactor fission products are very sensitive to power level. The radioactivity is calculated with a developed formula based on fuel basket loading integrity. (author)

Mechanism of deposit formation on fuel-wetted metal surfaces

Energy Technology Data Exchange (ETDEWEB)

Stavinoha, L.L.; Westbrook, S.R.; McInnis, L.A. [Southwest Research Institute, San Antonio, TX (United States)

1995-05-01

Experiments were performed in a Single-Tube Heat Exchanger (STHE) apparatus and a Hot Liquid Process Simulator (HLPS) configured and operated to meet Jet Fuel Thermal Oxidation Tester (JFTOT) ASTM D 3241 requirements. The HLPS-JFTOT heater tubes used were 1018 mild steel, 316 stainless steel (SS), 304 stainless steel (SS), and 304 SS tubes coated with aluminum, magnesium, gold, and copper. A low-sulfur Jet A fuel with a breakpoint temperature of 254{degrees}C was used to create deposits on the heater tubes at temperatures of 300{degrees}C, 340{degrees}C, and 380{degrees}C. Deposit thickness was measured by dielectric breakdown voltage and Auger ion milling. Pronounced differences between the deposit thickness measuring techniques suggested that both the Auger milling rate and the dielectric strength of the deposit may be affected by deposit morphology/composition (such as metal ions that may have become included in the bulk of the deposit). Carbon burnoff data were obtained as a means of judging the validity of DMD-derived deposit evaluations. ESCA data suggest that the thinnest deposit was on the magnesium-coated test tube. The Scanning Electron Microscope (SEM) photographs showed marked variations in the deposit morphology and the results suggested that surface composition has a significant effect on the mechanism of deposition. The most dramatic effect observed was that the bulk of deposits moved to tube locations of lower temperature as the maximum temperature of the tube was increased from 300{degrees} to 380{degrees}C, also verified in a single-tube heat exchanger. The results indicate that the deposition rate and quantity at elevated temperatures is not completely temperature dependent, but is limited by the concentration of dissolved oxygen and/or reactive components in the fuel over a temperature range.

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

Energy Technology Data Exchange (ETDEWEB)

McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ross, Steven [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Grey, Carissa Ann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arviso, Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Garmendia, Rafael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Fernandez Perez, Ismael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Palacio, Alejandro [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Calleja, Guillermo [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Garrido, David [COORDINADORA, Madrid (Spain); Rodriguez Casas, Ana [COORDINADORA, Madrid (Spain); Gonzalez Garcia, Luis [COORDINADORA, Madrid (Spain); Chilton, Lyman Wes [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Walz, Jacob [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gershon, Sabina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klymyshyn, Nicholas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pena, Ruben [Transportation Technology Center, Inc., Pueblo, CO (United States); Walker, Russell [Transportation Technology Center, Inc., Pueblo, CO (United States)

2018-01-01

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

International conference on storage of spent fuel from power reactors. Book of extended synopses

International Nuclear Information System (INIS)

2003-01-01

, high burnup fuel, radiation and heat load of high burnup and MOX fuel, advanced materials for storage racks and baskets, quality and maintenance of records

Fuel assembly

International Nuclear Information System (INIS)

Fujibayashi, Toru.

1970-01-01

Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

STS-95 crew members Duque and Mukai check out slidewire basket

Science.gov (United States)

1998-01-01

At Launch Pad 39-B, STS-95 Mission Specialist Pedro Duque of Spain (left) and Payload Specialist Chiaki Mukai look over the gate for the slidewire basket, part of the emergency egress system on the pad. Mukai represents the National Space Development Agency of Japan (NASDA), and Duque the European Space Agency (ESA). The STS-95 crew are at KSC to participate in a Terminal Countdown Demonstration Test (TCDT) which includes mission familiarization activities, emergency egress training, and a simulated main engine cut-off exercise. Other STS-95 crew members are Mission Specialist Stephen K. Robinson, Mission Commander Curtis L. Brown, Pilot Steven W. Lindsey, Payload Specialists John H. Glenn Jr., senator from Ohio, and Mission Specialist Scott E. Parazynski. The STS-95 mission, targeted for liftoff on Oct. 29, includes research payloads such as the Spartan solar-observing deployable spacecraft, the Hubble Space Telescope Orbital Systems Test Platform, the International Extreme Ultraviolet Hitchhiker, as well as the SPACEHAB single module with experiments on space flight and the aging process. Following the TCDT, the crew will be returning to Houston for final flight preparations.

Fuel can for a nuclear reactor

International Nuclear Information System (INIS)

Shimizu, Shigeo.

1984-01-01

Purpose: To decrease the possibility of damages in a fuel can by avoiding the close contact of the outer circumferential surface of a pellet to the entire inner circumference of the fuel can in the case if the pellet undergoes heat expansion. Constitution: The inner circumference of a fuel can includes at least three linear portions each with an equi-angular distance. The center for the circle (radius R2) inscribing each of the linear portions aligns with the axial center of the fuel can. A gap is formed to each inscribing circle with a band-like circular inner wall. The radius R2 for the inscribing circle is made larger than the radius R1 for the pellet and the length of the linear portion and the radius R2 for the inscribing circle are determined to desired values in view of the fuel design. If the fuel pellet expands thermally during reactor operation, since a gap is remained between the outer circumferential surface of the pellet and the inner circumferential surface of the fuel can and the outer circumferential surface of the pellet is not in close contact entirely with the inner circumferential surface of the fuel can, the possibility of damaging the fuel can is decreased. (Seki, T.)

Study on leakage rates of high temperature water from wet-type transport casks for spent fuel. Pt. 2. Leakage rates from a scratch on O-ring surface and narrow wires adhering to O-ring surface

International Nuclear Information System (INIS)

Asano, R.; Aritomi, M.; Sudi, A.; Kohketsu, Y.

1997-01-01

A programme for enhancement of fuel burnup has been promoted in Japan as part of the sophisticated programme for light water reactors to reduce the fuel cost and the amount of spent fuel. As part of this fuel programme, a new wet-type transport cask has been developed to transport the high burnup fuels efficiently. The purpose of this work is to clarify the margin of safety in the evaluation of the release rate of radioactive materials from the wet-type transport cask into the environment and to establish a practical evaluation method for leakage rates on leak behaviour of high temperature water from the casks. In this paper, leakage rates of water under high pressures and at high temperatures are investigated from two kinds of leak path model. One is a disc with a scratch on the surface which simulates a defect on the seal surface of the O-ring flange and the other is narrow stainless steel wires installed on the O-ring surface which simulates hair adhering to the O-ring surface. From the results, an evaluation method for the leakage rate of water under high pressure and at high temperature from a non-circular leak path and multiple leak paths is proposed. (author)

Reformulated and alternative fuels: modeled impacts on regional air quality with special emphasis on surface ozone concentration.

Science.gov (United States)

Schell, Benedikt; Ackermann, Ingmar J; Hass, Heinz

2002-07-15

The comprehensive European Air Pollution and Dispersion model system was used to estimate the impacts of the usage of reformulated and alternative fuels on regional air quality with special emphasis on surface ozone concentrations. A severe western European summer smog episode in July 1994 has been used as a reference, and the model predictions have been evaluated for this episode. A forecast simulation for the year 2005 (TREND) has been performed, including the future emission development based on the current legislation and technologies available. The results of the scenario TREND are used as a baseline for the other 2005 fuel scenarios, including fuel reformulation, fuel sulfur content, and compressed natural gas (CNG) as an alternative fuel. Compared to the year 1994, significant reductions in episode peak ozone concentrations and ozone grid hours are predicted for the TREND scenario. These reductions are even more pronounced within the investigated alternative and reformulated fuel scenarios. Especially, low sulfur fuels are appropriate for an immediate improvement in air quality, because they effect the emissions of the whole fleet. Furthermore, the simulation results indicate that the introduction of CNG vehicles would also enhance air quality with respect to ozone.

A method of surface area measurement of fuel materials by fission gas release at low temperature

International Nuclear Information System (INIS)

Kaimal, K.N.G.; Naik, M.C.; Paul, A.R.; Venkateswarlu, K.S.

1989-01-01

The present report deals with the development of a method for surface area measurement of nuclear fuel as well as fissile doped materials by fission gas release study at low temperature. The method is based on the evaluation of knock-out release rate of fission 133 Xe from irradiated fuel after sufficient cooling to decay the short lived activity. The report also describes the fabrication of an ampoule breaker unit for such study. Knock-out release rate of 133 Xe has been studied from UO 2 powders having varying surface area 'S' ranging from 270 cm 2 /gm to 4100 cm 2 /gm at two fissioning rates 10 12 f/cm 3 . sec. and 3.2x10 10 f/cm.sec. A relation between K and A has been established and discussed in this report. (author). 6 refs

Fuel pellet

International Nuclear Information System (INIS)

Hayashi, K.

1980-01-01

Fuel pellet for insertion into a cladding tube in order to form a fuel element or a fuel rod. The fuel pellet has got a belt-like projection around its essentially cylindrical lateral circumferential surface. The upper and lower edges in vertical direction of this belt-like projection are wave-shaped. The projection is made of the same material as the bulk pellet. Both are made in one piece. (orig.) [de

Ignition probability of fine dead surface fuels in native Patagonia forests of Argentina

Energy Technology Data Exchange (ETDEWEB)

Bianchi, L.; Defosse, G. E.

2014-06-01

Aim of study: The Canadian Forest Fire Weather Index (FWI) is being implemented all over the world. This index is being adapted to the Argentinean ecosystems since the year 2000. With the objective of calibrating the Fine Fuel Moisture Code (FFMC) of the FWI system to Patagonian forests, we studied the relationship between ignition probability and fine dead surface fuel moisture content (MC) as an indicator of potential fire ignition. Area of study: The study area is located in northwestern Patagonia, Argentina, and comprised two main forest types (cypress and nire) grown under a Mediterranean climate, with a dry summer and precipitations during winter and autumn ({approx}500-800 mm per year). Material and methods: We conducted lab ignition tests fires to determine the threshold of fine dead fuel ignition at different MC levels. Moisture content of dead fine surface fuels in the field was measured every 10-15 days from November to March for three seasons. We calculated the FFMC during these seasons and correlated it with the measured MC by applying a logistic regression model. We combined the results of the ignition tests and of the regressions to suggest FFMC categories for estimating fire danger in Patagonian forests. Main results: The ignition threshold occurred at MC values of 21.5 and 25.0% for cypress and nire sites, respectively. The MC measured varied from 7.3 to 129.6%, and the calculated FFMC varied between 13.4 and 92.6. Highly significant regressions resulted when FFMC was related to MC. The ignition threshold corresponded to a FFMC = 85. We proposed to divide the FFMC scale in three fire danger categories: Low (FFMC {<=} 85), High (85 < FFMC{<=}89) and Extreme (FFMC > 89). Research highlights: Our results provide a useful tool for predicting fire danger in these ecosystems, and are a contribution to the development of the Argentinean Fire Danger Rating and a reference for similar studies in other countries where the FWI is being implemented. (Author)

Biophysical Mechanistic Modelling Quantifies the Effects of Plant Traits on Fire Severity: Species, Not Surface Fuel Loads, Determine Flame Dimensions in Eucalypt Forests.

Science.gov (United States)

Zylstra, Philip; Bradstock, Ross A; Bedward, Michael; Penman, Trent D; Doherty, Michael D; Weber, Rodney O; Gill, A Malcolm; Cary, Geoffrey J

2016-01-01

The influence of plant traits on forest fire behaviour has evolutionary, ecological and management implications, but is poorly understood and frequently discounted. We use a process model to quantify that influence and provide validation in a diverse range of eucalypt forests burnt under varying conditions. Measured height of consumption was compared to heights predicted using a surface fuel fire behaviour model, then key aspects of our model were sequentially added to this with and without species-specific information. Our fully specified model had a mean absolute error 3.8 times smaller than the otherwise identical surface fuel model (p fire severity are the species of plants present rather than the surface fuel load, and demonstrate the accuracy and versatility of the model for quantifying this.

Fuel assembly and fuel cladding tube

International Nuclear Information System (INIS)

Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

1996-01-01

A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

Testing external surface of fuel element tubes for power nuclear reactors

International Nuclear Information System (INIS)

Naugol'nykh, O.G.; Nelyubin, Yu.V.

1987-01-01

Optical methods are regarded perspective for discovery and detection of flaws of external surfaces of fuel element tubes. The TV method has highest information content among them. Two mock-ups of facilities based on the TV method using a ''dissector'' type TV device and a TV tube with charge accumulation (vidikon) have been developed. It is concluded that complex testing - combination of ultrasonic, photoelectric and TV methods in a facility is necessary for discovery and analysis of the whole variety of flaws, though sensitivity of the TV method is enough for disclosure of all the main defects

A Continuous Liquid-Level Sensor for Fuel Tanks Based on Surface Plasmon Resonance

Directory of Open Access Journals (Sweden)

Antonio M. Pozo

2016-05-01

Full Text Available A standard problem in large tanks at oil refineries and petrol stations is that water and fuel usually occupy the same tank. This is undesirable and causes problems such as corrosion in the tanks. Normally, the water level in tanks is unknown, with the problems that this entails. We propose herein a method based on surface plasmon resonance (SPR to detect in real time the interfaces in a tank which can simultaneously contain water, gasoline (or diesel and air. The plasmonic sensor is composed of a hemispherical glass prism, a magnesium fluoride layer, and a gold layer. We have optimized the structural parameters of the sensor from the theoretical modeling of the reflectance curve. The sensor detects water-fuel and fuel-air interfaces and measures the level of each liquid in real time. This sensor is recommended for inflammable liquids because inside the tank there are no electrical or electronic signals which could cause explosions. The sensor proposed has a sensitivity of between 1.2 and 3.5 RIU−1 and a resolution of between 5.7 × 10−4 and 16.5 × 10−4 RIU.

Nuclear fuel element

International Nuclear Information System (INIS)

Mogard, J.H.

1977-01-01

A nuclear fuel element is disclosed for use in power producing nuclear reactors, comprising a plurality of axially aligned ceramic cylindrical fuel bodies of the sintered type, and a cladding tube of metal or metal alloys, wherein said cladding tube on its cylindrical inner surface is provided with a plurality of slightly protruding spacing elements distributed over said inner surface

Shipments of irradiated DIDO fuel from Risoe National Laboratory to the Savannah River Site - Challenges and achievements

International Nuclear Information System (INIS)

Anne, C.; Patterson, J.

2003-01-01

On September 28, 2000, the Board of Governors of Risoe National Laboratory decided to shut down the Danish research reactor DR3 due to technical problems (corrosion on the reactor aluminum tank). Shortly thereafter, the Danish Government asked the National Laboratory to empty the reactor and its storage pools containing a total of 255 DIDO irradiated elements and ship them to Savannah River Site in the USA as soon as possible. Risoe National Laboratory had previously contracted with Cogema Logistics to ship DR3 DIDO fuel elements to SRS through the end of the return program. The quantity of fuel was less than originally intended but the schedule was significantly shorter. It was agreed in June 2001 that a combination of Cogema Logistics' and NAC casks would be preferable, as it would allow Risoe to ship all the irradiated fuel in two shipments and complete the shipments by June 2002. Risoe National Laboratory, Cogema Logistics and NAC International had twelve months to perform the shipments including licensing, basket fabrication for the NAC-LWT casks and actual transport. The paper describes the challenging work that was accomplished to meet the date of June 2002. (author)

Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

International Nuclear Information System (INIS)

Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

2004-01-01

Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows f or surface

Dry Cask Storage Characterization Project - Phase 1: CASTOR V/21 Cask Opening and Examination

Energy Technology Data Exchange (ETDEWEB)

Bare, Walter Claude; Ebner, Matthias Anthony; Torgerson, Laurence Dale

2001-08-01

This report documents visual examination and testing conducted in 1999 and early 2000 at the Idaho National Engineering and Environmental Laboratory (INEEL) on a Gesellschaft für Nuklear Service (GNS) CASTOR V/21 pressurized water reactor (PWR) spent fuel dry storage cask. The purpose of the examination and testing is to develop a technical basis for renewal of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level waste at independent spent fuel storage installation sites. The examination and testing was conducted to assess the condition of the cask internal and external surfaces, cask contents consisting of 21 Westinghouse PWR spent fuel assemblies from Dominion’s (formerly named Virginia Power) Surry Power Station and cask concrete storage pad. The assemblies have been continuously stored in the CASTOR cask since 1985. Cask exterior surface and selected fuel assembly temperatures, and cask surface gamma and neutron dose rates were measured. Cask external/internal surfaces, fuel basket components including accessible weldments, fuel assembly exteriors, and primary lid seals were visually examined. Selected fuel rods were removed from one fuel assembly, visually examined, and then shipped to Argonne National Laboratory for nondestructive, destructive, and mechanical examination. Cask interior crud samples and helium cover gas samples were collected and analyzed. The results of the examination and testing indicate the concrete storage pad, CASTOR V/21 cask, and cask contents exhibited sound structural and seal integrity and that long-term storage has not caused detectable degradation of the spent fuel cladding or the release of gaseous fission products between 1985 and 1999.

Solar drying of West Indian ginger (Zingiber officinale Roscoe) rhizome using a wire basket dryer

Energy Technology Data Exchange (ETDEWEB)

Balladin, D.A.; Headley, O. [University of the West Indies (Barbados). Dept. of Chemistry; Chang Yen, I. [University of the West Indies, St. Augustine (Trinidad and Tobago). Dept. of Chemistry; McGaw, D.R. [University of the West Indies, St. Augustine (Trinidad and Tobago). Dept. of Chemical Engineering

1996-09-01

A wire basket dryer (1.8 m x 0.9 m x 0.2 m) was used to dry sliced (0.15 cm) West Indian ginger (Zingiber officinale Roscoe) rhizome to an acceptable moisture content of 10.2% (dry weight basis) over a 3 day period. The optimum charge size was 14.97 kg, with a packing density of 462.04 kg m{sup -3} and a specific drying rate of 0.446 h{sup -1}. The quantities (determined by high-pressure liquid chromatography) of the main pungent principles (ginerol and shogaol) extracted from fresh, non-steam-distilled solar-dried and steam-distilled solar-dried ginger rhizomes showed increases of 0.068, 0.46 and 0.67 g [per 100 g (dry weight basis)], respectively, with a decrease in the oleoresin quality (reflected in pungency profile) of the same order. (author)

Hot Surface Ignition

OpenAIRE

Tursyn, Yerbatyr; Goyal, Vikrant; Benhidjeb-Carayon, Alicia; Simmons, Richard; Meyer, Scott; Gore, Jay P.

2015-01-01

Undesirable hot surface ignition of flammable liquids is one of the hazards in ground and air transportation vehicles, which primarily occurs in the engine compartment. In order to evaluate the safety and sustainability of candidate replacement fuels with respect to hot surface ignition, a baseline low lead fuel (Avgas 100 LL) and four experimental unleaded aviation fuels recommended for reciprocating aviation engines were considered. In addition, hot surface ignition properties of the gas tu...

Hardened over-coating fuel particle and manufacture of nuclear fuel using its fuel particle

International Nuclear Information System (INIS)

Yoshimuda, Hideharu.

1990-01-01

Coated-fuel particles comprise a coating layer formed by coating ceramics such as silicon carbide or zirconium carbide and carbons, etc. to a fuel core made of nuclear fuel materials. The fuel core generally includes oxide particles such as uranium, thorium and plutonium, having 400 to 600 μm of average grain size. The average grain size of the coated-fuel particle is usually from 800 to 900 μm. The thickness of the coating layer is usually from 150 to 250 μm. Matrix material comprising a powdery graphite and a thermosetting resin such as phenol resin, etc. is overcoated to the surface of the coated-fuel particle and hardened under heating to form a hardened overcoating layer to the coated-fuel particle. If such coated-fuel particles are used, cracks, etc. are less caused to the coating layer of the coated-fuel particles upon production, thereby enabling to prevent the damages to the coating layer. (T.M.)

Canisters and nonfuel components at commercial nuclear reactors

International Nuclear Information System (INIS)

Gibbard, K.; Disbrow, J.

1994-01-01

This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

Surface contamination of spent fuel convoys - resumption of transport in France

International Nuclear Information System (INIS)

Pertuis, V.

2000-01-01

In France, 1998 was marked by the transport of spent fuel from EDF plants being suspended and then resumed. From the time the first inspections were carried out by the Nuclear Installations Safety Directorate (NISD), in charge of monitoring radioactive and fissile material for civil use since June 1997, surface contamination was found in a high percentage of packages and/or wagons containing spent fuel. The different expert appraisals showed that this had no consequences for the health of the public or of workers. Aiming at the resumption of transport, EDF and Cogema presented to the safety authority a plan of action including an increase in monitoring (number of points and cross-checking by SGS Qualitest), more widespread observance of good practices resulting from analyses by EDF and conclusions of its nuclear inspectorate, and an improvement in radiological cleanliness in the area where casks were loaded. During the inspections carried out at EDF plants, the NISD verified the application of this plan. Several observations were, nevertheless, made regarding maintenance of equipment, failure to apply procedures on a corporate level and the traceability of certain operations. The measures taken to sufficiently inform the public were applied. The NISD is continuing its monitoring actions to ensure that all EDF plants adopt best practices. However, the overall clean-up of EDF plants is a long-term operation. Finally, the NISD is continuing its monitoring of the different stages of spent fuel transport as well as other types of transport of radioactive materials associated with nuclear activities. (author)

Weigh Station and Grid Plate Testing

International Nuclear Information System (INIS)

PAJUNEN, A.L.

2000-01-01

The purpose of this test is to verify that the Shortened Fuel Canister Hook with Certified Scale (i.e. Weigh Station) can be used to weigh an empty canister from the Canister Well and the empty Primary Cleaning Machine (PCM) Strainer Basket from the process table. Drawing H-1-84835, ''Canister Handling Hook for Fuel Retrieval System Process Table,'' provides details of the Shortened Fuel Canister Hook. It is also necessary to verify that the grid plate can be lifted and tilted over a canister in the canister well. This testing shall be performed before N Reactor fuel is processed through the FRS in Phase 3. The Phase 3 Test will repeatedly weigh fuel and scrap canisters and the PCM strainer basket containing N Reactor fuel (Pajunen, et. al, 2000). Advance testing of this weigh station will ensure that accurate fuel weight data can be recorded in the Phase 3 Test. This document satisfies the requirements EN-6-031-00, ''Testing Process'' for a test plan, test specification and test procedure

Spent-fuel special-studies progress report: probable mechanisms for oxidation and dissolution of single-crystal UO2 surfaces

International Nuclear Information System (INIS)

Wang, R.

1981-03-01

Due to the complexity of the structural, microstructural and compositional characteristics of spent fuel, basic leaching and dissolution mechanisms were studied with UO 2 matrix material, specifically with single-crystal UO 2 , to isolate individual contributory factors. The effects of oxidation and oxidation-dissolution were investigated in different oxidation conditions, such as in air, oxygenated solutions and deionized water containing H 2 O 2 . In addition, the effects of temperature on dissolution of UO 2 were studied in autoclaves at 75 and 150 0 C. Also, oxidation and dissolution measurements were investigated via electrochemical methods to determine if those techniques could be applied to the characterization of leaching and dissolution of spent fuel in a hot cell. Finally, the effects of radiation were explored since the radiolysis of water may create a localized oxidizing condition at or near the spent fuel-solution interface, even in neutral or reducing conditions as commonly found in deep geological environments. The oxidation and oxidation-dissolution mechanisms for UO 2 are proposed as follows: The UO 2 surface is first oxidized in solution to form a UO/sub 2+x/ surface layer several angstroms thick. This oxidized surface has a high dissolution rate since the UO/sub 2+x/ reacts with the dissolved O 2 , or H 2 O 2 , to form uranyl complex ions in a U(VI) state. As the uranyl ions exceed the solubility limits in solution, they become hydrolyzed to form solid deposits and suspended particles of UO 3 hydrates. The thickness and porosity of the deposited UO 3 hydrate surface-film is dependent on temperature, pH and deposition time. A long-term dissolution rate is then determined by the nature of the surface film, such as porosity, solubility and mechanical properties

Putting all your eggs in one basket: life-history strategies, bet hedging, and diversification.

Science.gov (United States)

White, Andrew Edward; Li, Yexin Jessica; Griskevicius, Vladas; Neuberg, Steven L; Kenrick, Douglas T

2013-05-01

Diversification of resources is a strategy found everywhere from the level of microorganisms to that of giant Wall Street investment firms. We examine the functional nature of diversification using life-history theory-a framework for understanding how organisms navigate resource-allocation trade-offs. This framework suggests that diversification may be adaptive or maladaptive depending on one's life-history strategy and that these differences should be observed under conditions of threat. In three studies, we found that cues of mortality threat interact with one index of life-history strategy, childhood socioeconomic status (SES), to affect diversification. Among those from low-SES backgrounds, mortality threat increased preferences for diversification. However, among those from high-SES backgrounds, mortality threat had the opposite effect, inclining people to put all their eggs in one basket. The same interaction pattern emerged with a potential biomarker of life-history strategy, oxidative stress. These findings highlight when, and for whom, different diversification strategies can be advantageous.

Combustor and method for distributing fuel in the combustor

Science.gov (United States)

Uhm, Jong Ho; Ziminsky, Willy Steve; Johnson, Thomas Edward; York, William David

2016-04-26

A combustor includes a tube bundle that extends radially across at least a portion of the combustor. The tube bundle includes an upstream surface axially separated from a downstream surface. A plurality of tubes extends from the upstream surface through the downstream surface, and each tube provides fluid communication through the tube bundle. A baffle extends axially inside the tube bundle between adjacent tubes. A method for distributing fuel in a combustor includes flowing a fuel into a fuel plenum defined at least in part by an upstream surface, a downstream surface, a shroud, and a plurality of tubes that extend from the upstream surface to the downstream surface. The method further includes impinging the fuel against a baffle that extends axially inside the fuel plenum between adjacent tubes.

Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

1984-10-01

Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel

International Nuclear Information System (INIS)

Yoo, Seung Hun

2011-02-01

parameters including grid sizes, radiation model, boundary conditions, a small-scale model was designed and tested which has the main components of a dry storage cask, such as fuel rods, basket and cask. Based on the sensitivity study the full-scope simulations by FLUENT were performed and compared with the experiment data and COBRA-SFS results. FLUENT results for the perdition of temperature distribution were agreed within measurement uncertainty. However, there were larger deviations for prediction of velocity between FLUENT and COBRA-SFS. The velocity predicted by COBRA-SFS was much higher than that predicted by FLUENT. The sensitivity studies with the full-scope simulation model were performed for a grid size and basket gap sizes. It was revealed that the basket gap size was the most sensitive parameter deciding the PCT in the dry storage cask. By this study, we proposed an applicable guideline of CFD use for the full-scope simulation of the dry storage cask. In the third study, a new type of the interim storage, which consists of one tunnel with eighty metal storage casks and one chimney, was proposed and studied. The tunnel-type storage has the following advantages: the reduction of the risk of the air plane crash and the easy control of the capacity of the system through the height of the chimney. The thermal-hydraulic characteristics of the tunnel storage were analyzed by FLUENT. The CFD simulation methodology was validated against the experimental data of a vault storage system (Sakamoto et al., 2000) because of the similar thermal-hydraulic characteristics between the tunnel and the vault. Then, the sensitivity studies of three parameters were done: the tunnel geometry, the cask arrangement and the turbulence. The local peak temperature on the metal casks is mainly governed by mass flow rates in the gaps. V1 case was recommended for the accommodation of the metal storage casks in the tunnel. The staggered cask arrangement was recommended due to the advantageous

Characterizing hand-piled fuels

Science.gov (United States)

Clinton S. Wright; Paige C. Eagle; Cameron S. Balog

2010-01-01

Land managers throughout the West pile and burn surface fuels to mitigate fire hazard in dry forests. Whereas piling was historically conducted with heavy machinery following commercial harvesting operations, land managers are increasingly prescribing the use of hand piling and burning to treat surface fuels created by thinning and brush cutting. An estimate of the...

Crud deposition on fuel in WWER reactors

International Nuclear Information System (INIS)

Kysela, J.; Svarc, V.; Androva, K.; Ruzickova, M.

2008-01-01

Reliability of nuclear fuel and radiation fields surrounding primary systems are important aspects of overall nuclear reactor safety. Corrosion product (crud) deposition on fuel surfaces has implications for fuel performance through heat transfer and local chemistry modifications. Crud is currently one of the key industry issues and has been implicated in several recent cases of crud-related fuel failures and core plugging. Activated crud is deposited on out-of-core surfaces, mainly steam generators, resulting in high radiation fields and high doses of plant staff. Due to radiation build-up in primary circuit systems, decontamination of primary systems components and steam generators is used. Several issues involving decontamination were observed in some cases. After decontamination higher corrosion product release occurs followed by subsequent crud deposition on fuel surfaces. The paper summarizes experience with water chemistry and decontamination that can influence crud deposition on fuel surfaces. The following areas are discussed: 1) Experience with crud deposition, primary water chemistry and decontamination under operating conditions; 2) The behaviour of organic compounds in primary coolant and on fuel surfaces; 3) A proposed experimental programme to study crud deposition. (authors)

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.)

Measurements of Fundamental Fluid Physics of SNF Storage Canisters

Energy Technology Data Exchange (ETDEWEB)

Condie, Keith Glenn; Mc Creery, Glenn Ernest; McEligot, Donald Marinus

2001-09-01

With the University of Idaho, Ohio State University and Clarksean Associates, this research program has the long-term goal to develop reliable predictive techniques for the energy, mass and momentum transfer plus chemical reactions in drying / passivation (surface oxidation) operations in the transfer and storage of spent nuclear fuel (SNF) from wet to dry storage. Such techniques are needed to assist in design of future transfer and storage systems, prediction of the performance of existing and proposed systems and safety (re)evaluation of systems as necessary at later dates. Many fuel element geometries and configurations are accommodated in the storage of spent nuclear fuel. Consequently, there is no one generic fuel element / assembly, storage basket or canister and, therefore, no single generic fuel storage configuration. One can, however, identify generic flow phenomena or processes which may be present during drying or passivation in SNF canisters. The objective of the INEEL tasks was to obtain fundamental measurements of these flow processes in appropriate parameter ranges.

Measurement of dose rate and estimation of beta activity in zircaloy hull drum

International Nuclear Information System (INIS)

Pandey, J.P.N.; Kumar, Pankaj; Shinde, A.M.; Purohit, R.G.; Sarkar, P.K.

2012-01-01

Fuel Reprocessing Plant is designed for the processing of spent fuel from reactor for the recovery of plutonium and uranium as PuO 2 and U 3 O 8 respectively. Zircaloy is used as cladding material of natural uranium fuel pins used in the reactors. In reprocessing plants chop and leach method is used to remove the zircaloy clad from the fuel matrix during Head End Treatment. Initially spent fuel bundles are chopped into pieces and collected in perforated baskets kept in dissolvers. All chopped pieces are dissolved in HNO 3 in the dissolvers followed by heating and boiling. Dissolved solutions are transferred to Filtrate Tank (FT) leaving behind un-dissolved zircoloy hull pieces in the dissolver baskets. Un-dissolved and almost dry hull pieces are transferred in hull drum from the dissolver baskets using the Hull Tilting Facility. Hull drums are made of stainless steel having 500 litre capacity and two third of its volume is filled with zircoloy pieces. Hull drums filled with hull pieces are loaded in Hull Removal Cask (HRC) and transported to SWMF (Solid Waste Management Facility) site for interim storage/disposal in tile holes. Hull pieces are high active solid wastes which contain significant amount of fission products. Radiation levels on hull drums are in the range of few hundreds of mGy/h which has high potential of external hazards if not handled properly. Therefore hull drums are handled remotely in specially designed lead shielded cask

Criticality effects of longitudinal gaps in poison for storage/transport casks

International Nuclear Information System (INIS)

Wells, A.H.

1985-01-01

A series of criticality calculations was performed with the AMPX/KENO system to determine the sensitivity of the NAC S/T cask 31 assembly basket, which is optimized for a design-basis fuel enrichment of 3.7 wt % 235 U, to axial gaps in the boron neutron poison. The results of these calculations show that axial gaps in the boron cause no statistically detectable change in k/sub eff/ until a minimum gap size is reached. The minimum gap size to change k/sub eff/ is dependent on the basket segment length, because a longer segment length results in fewer gaps for a given active fuel length. Longer segment lengths are less sensitive to gaps in the neutron poison. A typical segment length of 12 to 18 in. is projected for a casting of aluminum/boron alloy, indicating that axial gaps in the neutron poison of 1 in. would be acceptable. This gap thickness is much greater than the intersegment gap produced by modern casting techniques. The investigation described here demonstrated that an axial gap in neutron poison is acceptable for basket castings of large storage/transport casks. A precedent for such gaps is the NLI-6502 cask, so a cask basket with intersegment gaps should be licensable

Nuclear fuel element and container

International Nuclear Information System (INIS)

Grubb, W.T.; King, L.H.

1981-01-01

The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

Safety performance of a near surface repository subject to a fuel burning

International Nuclear Information System (INIS)

Stefanini, Lorenzo; Frano, Rosa Lo; Forasassi, Giuseppe

2015-01-01

This study aims to investigate the performances of a near surface repository subject to fuel burning occurring simultaneously or subsequently to a large commercial aircraft impact. Specifically the thermal effects caused by a Boeing-747 crushing (considered like “beyond design basis accident”) are studied. An important part of this study is the analysis of the possible (thermo-mechanical) degradation effects, as dehydration, degasification, pressurization, etc. that the concrete may undergo, particularly in the case of prolonged fire, and of the resistance of structure itself in this condition. Conservative assumptions and restrictions have been made with regard to the fire scenario, the maximum temperature of which is calculated on the basis of the fuel airplane amount, the normal impact, the variation of the material properties along with the temperature as well the damaging phenomena of concrete. The airplane impact load, calculated with the Riera approach, and the maximum temperature, reached during the fuel combustion, are used as input (boundary condition) in the numerical simulations performed by MARC© code. The obtained results showed that a repository wall thickness, ranging from 0.6 to 0.9 m, is not sufficient to prevent the local penetration of wall. To reduce the computational cost, the analyses have been made only on a half part of the structure, highlighting the dominance of thermal effects. Despite the ongoing concrete degradation phenomena, the overall integrity of the repository seemed to be guaranteed as well as the containment and the confinement of radioactive waste. (author)

Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Mohammadi, Ashgar; Omidvari, Nima [Iran Radioactive Waste Management Company, Tehran (Iran, Islamic Republic of); Hassanzadeh, Mostafa [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

2017-12-15

Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k{sub eff}) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

International Nuclear Information System (INIS)

Mohammadi, Ashgar; Omidvari, Nima; Hassanzadeh, Mostafa

2017-01-01

Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k eff ) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

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

Flow visualization of forced and natural convection in internal cavities. 1998 annual progress report

International Nuclear Information System (INIS)

Brodkey, R.S.; Clarksean, R.; Crepeau, J.C.; Guezennec, Y.G.; McEligot, D.M.

1998-01-01

'The objective of this research program is to understand the fluid physics when corroded spent nuclear fuel (SNF) elements are passivated by injecting treatment gases into a storage canister. By developing a reliable predictive technique for the energy, mass, and momentum transfer in the presence of surface reactions, transfer and storage systems can be efficiently and safely designed. The objective will be reached by using innovative flow visualization techniques and experimental measurements of the flow field to support computational models. This report summarizes work completed after eight months of a three-year, collaborative project. A generic idealization of a combined drying and passivation approach has been defined, which represents a section of a vertical canister with baskets of SNF elements. This simulation includes flow phenomena that occur in canisters for high- and/or low-enrichment fuels. A steady flow of the passivation fluid is introduced at the bottom of the canister via a central tube from the top. Fluid flows through an array of holes in the perforated basket support plate then around the simulated elements and out the top. Dimensions and flow rates for the idealized situation correspond to those for typical drying canisters. Approximate calculations have identified the ranges of values of flow parameters needed to determine the flow regimes occurring in practice.'

The concrete canister program

International Nuclear Information System (INIS)

Ohta, M.M.

1978-02-01

In the spring of 1974, WNRE began development and demonstration of a dry storage concept, called the concrete canister, as a possible alternative to storage of irradiated CANDU fuel in water pools. The canister is a thick-walled concrete monolith containing baskets of fuel in the dry state. The decay heat from the fuel is dissipated to the environment by natural heat transfer. Four canisters were designed and constructed. Two canisters containing electric heaters have been subjected to heat loads of 2.5 times the design, ramp heat-load cycling, and simulated weathering tests. The other two canisters were loaded with irradiated fuel, one containing fuel bundles of uniform decay heat and the other containing bundles of non-uniform decay heat in a non-symmetrical radial and axial array. The collected data were used to verify the analytical tools for prediction of effectiveness of heat transfer and radiation shielding and to verify the design of the basket and canisters. The demonstration canisters have shown that this concept is a viable alternative to water pools for the storage of irradiated CANDU fuel. (author)

A risk-informed basis for establishing non-fixed surface contamination limits for spent fuel transportation casks

International Nuclear Information System (INIS)

Rawl, R.R.; Eckerman, K.F.; Bogard, J.S.; Cook, J.R.

2004-01-01

The current limits for non-fixed contamination on packages used to transport radioactive materials were introduced in the 1961 edition of the International Atomic Energy Agency (IAEA) transport regulations and were based on radiation protection guidance and practices in use at that time. The limits were based on exposure scenarios leading to intakes of radionuclides by inhalation and external irradiation of the hands. These considerations are collectively referred to as the Fairbairn model. Although formulated over 40 years ago, the model remains unchanged and is still the basis of current regulatory-derived limits on package non-fixed surface contamination. There can also be doses that while not resulting directly from the contamination, are strongly influenced by and attributable to transport regulatory requirements for contamination control. For example, actions necessary to comply with the current derived limits for light-water-reactor (LWR) spent nuclear fuel (SNF) casks can result in significant external doses to workers. This is due to the relatively high radiation levels around the loaded casks, where workers must function during the measurement of contamination levels and while decontaminating the cask. In order to optimize the total dose received due to compliance with cask contamination levels, it is necessary to take into account all the doses that vary as a result of the regulatory limit. Limits for non-fixed surface contamination on spent fuel casks should be established by using a model that considers and optimizes the appropriate exposure scenarios both in the workplace and in the public environment. A risk-informed approach is needed to ensure optimal use of personnel and material resources for SNF-based packaging operations. This paper is a summary of a study sponsored by the US Nuclear Regulatory Commission and performed by Oak Ridge National Laboratory that examined the dose implications for removable surface contamination limits on spent fuel

Surface strontium enrichment on highly active perovskites for oxygen electrocatalysis in solid oxide fuel cells

KAUST Repository

Crumlin, Ethan J.; Mutoro, Eva; Liu, Zhi; Grass, Michael E.; Biegalski, Michael D.; Lee, Yueh-Lin; Morgan, Dane; Christen, Hans M.; Bluhm, Hendrik; Shao-Horn, Yang

2012-01-01

Perovskite oxides have high catalytic activities for oxygen electrocatalysis competitive to platinum at elevated temperatures. However, little is known about the oxide surface chemistry that influences the activity near ambient oxygen partial pressures, which hampers the design of highly active catalysts for many clean-energy technologies such as solid oxide fuel cells. Using in situ synchrotron-based, ambient pressure X-ray photoelectron spectroscopy to study the surface chemistry changes, we show that the coverage of surface secondary phases on a (001)-oriented La 0.8Sr 0.2CoO 3-δ (LSC) film becomes smaller than that on an LSC powder pellet at elevated temperatures. In addition, strontium (Sr) in the perovskite structure enriches towards the film surface in contrast to the pellet having no detectable changes with increasing temperature. We propose that the ability to reduce surface secondary phases and develop Sr-enriched perovskite surfaces of the LSC film contributes to its enhanced activity for O 2 electrocatalysis relative to LSC powder-based electrodes. © 2012 The Royal Society of Chemistry.

Using field data to assess model predictions of surface and ground fuel consumption by wildfire in coniferous forests of California

Science.gov (United States)

Lydersen, Jamie M.; Collins, Brandon M.; Ewell, Carol M.; Reiner, Alicia L.; Fites, Jo Ann; Dow, Christopher B.; Gonzalez, Patrick; Saah, David S.; Battles, John J.

2014-03-01

Inventories of greenhouse gas (GHG) emissions from wildfire provide essential information to the state of California, USA, and other governments that have enacted emission reductions. Wildfires can release a substantial amount of GHGs and other compounds to the atmosphere, so recent increases in fire activity may be increasing GHG emissions. Quantifying wildfire emissions however can be difficult due to inherent variability in fuel loads and consumption and a lack of field data of fuel consumption by wildfire. We compare a unique set of fuel data collected immediately before and after six wildfires in coniferous forests of California to fuel consumption predictions of the first-order fire effects model (FOFEM), based on two different available fuel characterizations. We found strong regional differences in the performance of different fuel characterizations, with FOFEM overestimating the fuel consumption to a greater extent in the Klamath Mountains than in the Sierra Nevada. Inaccurate fuel load inputs caused the largest differences between predicted and observed fuel consumption. Fuel classifications tended to overestimate duff load and underestimate litter load, leading to differences in predicted emissions for some pollutants. When considering total ground and surface fuels, modeled consumption was fairly accurate on average, although the range of error in estimates of plot level consumption was very large. These results highlight the importance of fuel load input to the accuracy of modeled fuel consumption and GHG emissions from wildfires in coniferous forests.

Potential impacts of crud deposits on fuel rod behaviour on high powered PWR fuel rods

International Nuclear Information System (INIS)

Wilson, W.; Comstock, R.J.

1999-01-01

Fuel assemblies operating with significant sub-cooled boiling are subject to deposition of surface deposits commonly referred to as crud. This crud can potentially cause concentration of chemical species within the deposits which can be detrimental to cladding performance in PWRs. In addition, these deposits on the surface of the cladding can result in power anomalies and erroneous reporting of fuel rod oxide thickness which can substantially hamper corrosion and core performance modeling efforts. Data is presented which illustrates the importance of accounting for the presence of crud on fuel cladding surfaces. Several methods used to correct for this phenomenon when collecting and analyzing zirconium alloy field oxide thickness measurements are described. Various observations related to crud characteristics and its impact on fuel rod performance are also addressed. (author)

Fuel removing method for high burnup fuel and device therefor

International Nuclear Information System (INIS)

Terakado, Shogo; Owada, Isao; Kanno, Yoshio; Aizawa, Sakue; Yamahara, Takeshi.

1993-01-01

A through hole is perforated at the center of a fuel rod in a cladding tube by a diamond drill in a water vessel. Further, the through hole is enlarged by the diamond drill. A pellet removing tool is attached to a drill chuck instead of the diamond drill. Then, the thin cylindrical fuel pellet remaining on the inner surface of the cladding tube is removed by using a pellet removing tool while applying vibrations. Subsequently, a wire brush having a slightly larger diameter than that of the inner diameter of the cladding tube is attached to the drill chuck and rotated to finish the inner surface, so that a small amount of pellets remained on the inner surface of the cladding tube is removed. Pellet powders in the water vessel are collected and recovered to the water container. This can remove high burnup fuels which are firmly sticked to the cladding tube, without giving thermal or mechanical influences on the cladding tube. (I.N.)

Methods of conditioning direct methanol fuel cells

Science.gov (United States)

Rice, Cynthia; Ren, Xiaoming; Gottesfeld, Shimshon

2005-11-08

Methods for conditioning the membrane electrode assembly of a direct methanol fuel cell ("DMFC") are disclosed. In a first method, an electrical current of polarity opposite to that used in a functioning direct methanol fuel cell is passed through the anode surface of the membrane electrode assembly. In a second method, methanol is supplied to an anode surface of the membrane electrode assembly, allowed to cross over the polymer electrolyte membrane of the membrane electrode assembly to a cathode surface of the membrane electrode assembly, and an electrical current of polarity opposite to that in a functioning direct methanol fuel cell is drawn through the membrane electrode assembly, wherein methanol is oxidized at the cathode surface of the membrane electrode assembly while the catalyst on the anode surface is reduced. Surface oxides on the direct methanol fuel cell anode catalyst of the membrane electrode assembly are thereby reduced.

Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Jegou, C., E-mail: christophe.jegou@cea.f [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Caraballo, R.; De Bonfils, J.; Broudic, V.; Peuget, S. [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Vercouter, T. [Commissariat a l' Energie Atomique (CEA), Saclay Reasearch Center, B.P. 11, F-91191 Gif-sur-Yvette Cedex (France); Roudil, D. [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France)

2010-04-01

The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2 (registered) ) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 deg. C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO{sub 2} matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO{sub 2} grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH){sub 4(am)} phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.

Fuel spacer

International Nuclear Information System (INIS)

Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

1992-01-01

The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

Modeling electrochemical resistance with coal surface properties in a direct carbon fuel cell based on molten carbonate

Science.gov (United States)

Eom, Seongyong; Ahn, Seongyool; Kang, Kijoong; Choi, Gyungmin

2017-12-01

In this study, a numerical model of activation and ohmic polarization is modified, taking into account the correlation function between surface properties and inner resistance. To investigate the correlation function, the surface properties of coal are changed by acid treatment, and the correlations between the inner resistance measured by half-cell tests and the surface characteristics are analyzed. A comparison between the model and experimental results demonstrates that the absolute average deviations for each fuel are less than 10%. The numerical results show that the sensitivities of the coal surface properties affecting polarization losses change depending on the operating temperature. The surface oxygen concentrations affect the activation polarization and the sensitivity decreased with increasing temperature. The surface ash of coal is an additional index to be considered along with ohmic polarization and it has the greatest effect on the surface properties at 973 K.

Applications to shielding design and others of monte carlo method

Energy Technology Data Exchange (ETDEWEB)

Ito, Daiichiro [Mitsui Engineering and Shipbuiding Co., Ltd., Tokyo (Japan)

2001-01-01

One-dimensional or two-dimensional Sn computer code (ANISN, DOT3.5, etc.) and a point attenuation kernel integral code (QAD, etc.) have been used widely for shielding design. Application examples of monte carlo method which could follow precisely the three-dimensional configuration of shielding structure are shown as follow: (1) CASTER cask has a complex structure which consists of a large number of fuel baskets (stainless steel), neutron moderators (polyethylene rods), the body (cast iron), and cooling fin. The R-{theta} model of Sn code DOT3.5 cannot follow closely the complex form of polyethylene rods and fuel baskets. A monte carlo code MORSE is used to ascertain the calculation results of DOT3.5. The discrepancy between the calculation results of DOT3.5 and MORSE was in 10% for dose rate at distance of 1 m from the cask surface. (2) The dose rates of an iron cell at 10 cm above the floor are calculated by the code QAD and the MORSE. The reflected components of gamma ray caused by the auxiliary floor shield (lead) are analyzed by the MORSE. (3) A monte carlo code MCNP4A is used for skyshine evaluation of spent fuel carrier ship 'ROKUEIMARU'. The direct and skyshine components of gamma ray and neutron flux are estimated at each center of engine room and wheel house. The skyshine dose rate of neutron flux is 5-15 times larger than the gamma ray. (M. Suetake)

Apparatus for mixing fuel in a gas turbine

Science.gov (United States)

Uhm, Jong Ho; Johnson, Thomas Edward

2015-04-21

A combustor nozzle includes an inlet surface and an outlet surface downstream from the inlet surface, wherein the outlet surface has an indented central portion. A plurality of fuel channels are arranged radially outward of the indented central portion, wherein the plurality of fuel channels extend through the outlet surface.

Classification of Steps on Road Surface Using Acceleration Signals

Directory of Open Access Journals (Sweden)

Junji Takahashi

2015-12-01

Full Text Available In order to reduce a road monitoring cost, we propose a system to monitor extensively road condition by cyclists with a smartphone. In this paper, we propose two methods towards road monitoring. First is to classify road signals to four road conditions. Second is to extract road signal from a smartphone's accelerometer in three positions: pants' side pocket, chest pocket and a bag in a front basket. In pants' side pocket, road signal is extracted by Independent Component Analysis. In chest pocket and bag in a front basket, road signal is extracted by selecting 1-axis affected from gravitational acceleration. In the experiment of the classification method, overall accuracy was 75%. The experimental results of the extraction methods with correlation coefficient showed the overall accuracy were more than 0.7 in pants' side pocket and chest pocket, the overall accuracy was less than 0.3 in bag in a front basket.

A food basket investigation during the autumn of 1994; Matkorgsundersoekning hoesten 1994

Energy Technology Data Exchange (ETDEWEB)

Moere, H.; Falk, R.; Svedjemark, G.A. [Swedish Radiation Protection Inst., Stockholm (Sweden); Becker, W.; Brugaard Konde, Aa. [National Food Administration, Uppsala (Sweden)

1995-10-01

During the autumn of 1994 an investigation of foodstuffs has been accomplished to assess the average intake of {sup 137}Cs by the Swedish population due to the Chernobyl accident. A standardized food basket has been collected from two grocers in 10 localities, of which the majority came from areas with the highest fallout. The estimated maximum intake of {sup 137}Cs was 815 Bq/year in the inland of the county of Vaesterbotten. The population weighted average intake for the fallout affected counties was 435 Bq/year. The rest of the county received an intake of 235 Bq/year. The population weighted average of the intake for the whole county was estimated to 274 Bq/year. From this intake the calculated body burden would be 1.3 Bq/kg for the average citizen. Whole-body measurements of a sample of the population gave 2.0 Bq/kg. A plausible explanation would be that 40% of the intake of {sup 137}Cs can have its origin from the 10% of the consumption of foodstuffs that are home produced or collected for the average individual in Sweden. The average intake of 274 Bq/year gives a committed effective dose equivalent of 3.6 {mu}Sv. 6 refs, 10 tabs.

A Refined Basket of Goods Approach for Comparing Construction Costs between Countries

Directory of Open Access Journals (Sweden)

Rick Best

2012-11-01

Full Text Available Most comparisons of construction industry performance requirethat construction costs be converted to a common base. Existingmechanisms for such conversions produce unreliable results.A proposed method for producing industry-specifi c conversionfactors was tested using a single building type. A basket ofmaterials and labour was identifi ed and weighted to refl ect the costshare of each item in a completed project. Prices for the basketwere gathered in three cities and simple construction specifi cconversion indices were calculated based on the constructionpurchasing power of each currency. The construction purchasingpower parities (CPPPs showed marked differences from otheravailable conversion mechanisms such as exchange rates andgeneral purchasing power parities (PPPs that have been used inprevious international comparison studies. While the study waslimited in scope, and is only the fi rst stage of a longer process, thesubstantial differences in comparative costs based on purchasingpower illustrate the problems inherent in international costcomparisons. For example, comparing Singapore and Sydney,Singapore costs appear to be only 40% of those in Sydney (basedon exchange rates about two-thirds the cost of Sydney (usinggeneral PPPs or almost the same (using the preliminary CPPPs.These results illustrate the problems of converting costs fromlocal currencies to a single base currency and suggest that furtherdevelopment is needed to improve the reliability of outcomes.

Deep geological disposal system development; mechanical structural stability analysis of spent nuclear fuel disposal canister under the internal/external pressure variation

Energy Technology Data Exchange (ETDEWEB)

Kwen, Y. J.; Kang, S. W.; Ha, Z. Y. [Hongik University, Seoul (Korea)

2001-04-01

This work constitutes a summary of the research and development work made for the design and dimensioning of the canister for nuclear fuel disposal. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for spent fuel disposal should be securred. Usually this repository is expected to locate at a depth of 500m underground. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for spent nuclear fuel disposal in a deep repository in the crystalline bedrock, which entails an evenly distributed load of hydrostatic pressure from undergroundwater and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables array type of inner baskets and thicknesses of outer shell and lid and bottom are tried to be determined through the mechanical linear structural analysis, thicknesses of outer shell is determined through the nonlinear structural analysis, and the bentonite buffer analysis for the rock movement is conducted through the of nonlinear structural analysis Also the thermal stress effect is computed for the cast iron insert. The canister types studied here are one for PWR fuel and another for CANDU fuel. 23 refs., 60 figs., 23 tabs. (Author)

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.)

Nanoporous silver cathode surface treated by atomic layer deposition of CeO_x for low-temperature solid oxide fuel cells

International Nuclear Information System (INIS)

Neoh, Ke Chean; Han, Gwon Deok; Kim, Manjin; Kim, Jun Woo; Choi, Hyung Jong; Park, Suk Won; Shim, Joon Hyung

2016-01-01

We evaluated the performance of solid oxide fuel cells (SOFCs) with a 50 nm thin silver (Ag) cathode surface treated with cerium oxide (CeO_x) by atomic layer deposition (ALD). The performances of bare and ALD-treated Ag cathodes were evaluated on gadolinia-doped ceria (GDC) electrolyte supporting cells with a platinum (Pt) anode over 300 °C–450 °C. Our work confirms that ALD CeO_x treatment enhances cathodic performance and thermal stability of the Ag cathode. The performance difference between cells using a Ag cathode optimally treated with an ALD CeO_x surface and a reference Pt cathode is about 50% at 450 °C in terms of fuel cell power output in our experiment. The bare Ag cathode completely agglomerated into islands during fuel cell operation at 450 °C, while the ALD CeO_x treatment effectively protects the porosity of the cathode. We also discuss the long-term stability of ALD CeO_x-treated Ag cathodes related to the microstructure of the layers. (paper)

Fuel assembly

International Nuclear Information System (INIS)

Nakatsuka, Masafumi; Matsuzuka, Ryuji.

1976-01-01

Object: To provide a fuel assembly which can decrease pressure loss of coolant to uniform temperature. Structure: A sectional area of a flow passage in the vicinity of an inner peripheral surface of a wrapper tube is limited over the entire length to prevent the temperature of a fuel element in the outermost peripheral portion from being excessively decreased to thereby flatten temperature distribution. To this end, a plurality of pincture-frame-like sheet metals constituting a spacer for supporting a fuel assembly, which has a plurality of fuel elements planted lengthwise and in given spaced relation within the wrapper tube, is disposed in longitudinal grooves and in stacked fashion to form a substantially honeycomb-like space in cross section. The fuel elements are inserted and supported in the space to form a fuel assembly. (Kamimura, M.)

Characteristics of used CANDU fuel relevant to the Canadian nuclear fuel waste management program

Energy Technology Data Exchange (ETDEWEB)

Wasywich, K M

1993-05-01

Literature data on the characteristics of used CANDU power reactor fuel that are relevant to its performance as a waste form have been compiled in a convenient handbook. Information about the quantities of used fuel generated, burnup, radionuclide inventories, fission gas release, void volume and surface area, fuel microstructure, fuel cladding properties, changes in fuel bundle properties due to immobilization processes, radiation fields, decay heat and future trends is presented for various CANDU fuel designs. (author). 199 refs., 39 tabs., 100 figs.

Characteristics of used CANDU fuel relevant to the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Wasywich, K.M.

1993-05-01

Literature data on the characteristics of used CANDU power reactor fuel that are relevant to its performance as a waste form have been compiled in a convenient handbook. Information about the quantities of used fuel generated, burnup, radionuclide inventories, fission gas release, void volume and surface area, fuel microstructure, fuel cladding properties, changes in fuel bundle properties due to immobilization processes, radiation fields, decay heat and future trends is presented for various CANDU fuel designs. (author). 199 refs., 39 tabs., 100 figs

Fuels planning: science synthesis and integration; forest structure and fire hazard fact sheet 05: fuel treatment principles for complex landscapes

Science.gov (United States)

Rocky Mountain Research Station USDA Forest Service

2004-01-01

Appropriate types of thinning and surface fuel treatments are clearly useful in reducing surface and crown fire hazards under a wide range of fuels and topographic situations. This paper provides well-established scientific principles and simulation tools that can be used to adjust fuel treatments to attain specific risk levels.

Rock cavern storage of spent fuel

Energy Technology Data Exchange (ETDEWEB)

Cho, Won Jin; Kim, Kyung Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Sang Ki [Inha University, Incheon (Korea, Republic of)

2015-12-15

The rock cavern storage for spent fuel has been assessed to apply in Korea with reviewing the state of the art of the technologies for surface storage and rock cavern storage of spent fuel. The technical feasibility and economic aspects of the rock cavern storage of spent fuel were also analyzed. A considerable area of flat land isolated from the exterior are needed to meet the requirement for the site of the surface storage facilities. It may, however, not be easy to secure such areas in the mountainous region of Korea. Instead, the spent fuel storage facilities constructed in the rock cavern moderate their demands for the suitable site. As a result, the rock cavern storage is a promising alternative for the storage of spent fuel in the aspect of natural and social environments. The rock cavern storage of spent fuel has several advantages compared with the surface storage, and there is no significant difference on the viewpoint of economy between the two alternatives. In addition, no great technical difficulties are present to apply the rock cavern storage technologies to the storage of domestic spent fuel.

Nuclear fuel element

International Nuclear Information System (INIS)

Iwano, Yoshihiko.

1993-01-01

Microfine cracks having a depth of less than 10% of a pipe thickness are disposed radially from a central axis each at an interval of less than 100 micron over the entire inner circumferential surface of a zirconium alloy fuel cladding tube. For manufacturing such a nuclear fuel element, the inside of the cladding tube is at first filled with an electrolyte solution of potassium chloride. Then, electrolysis is conducted using the cladding tube as an anode and the electrolyte solution as a cathode, and the inner surface of the cladding tube with a zirconium dioxide layer having a predetermined thickness. Subsequently, the cladding tube is laid on a smooth steel plate and lightly compressed by other smooth steel plate to form microfine cracks in the zirconium dioxide layer on the inner surface of the cladding tube. Such a compressing operation is continuously applied to the cladding tube while rotating the cladding tube. This can inhibit progress of cracks on the inner surface of the cladding tube, thereby enabling to prevent failure of the cladding tube even if a pellet/cladding tube mechanical interaction is applied. Accordingly, reliability of the nuclear fuel elements is improved. (I.N.)

Selection of the reference concept for the surface examination stations in the fuels and materials examination facility

International Nuclear Information System (INIS)

Frandsen, G.B.; Nash, C.R.

1978-01-01

The prototype surface examination station for the Fuels and Materials Examination Facility (FMEF) will use closed circuit television (CCTV) for routine modes of operation along with a nuclear periscope for special examination needs. The CCTV and the nuclear periscope were evaluated against prescribed station requirements and compared in a side-by-side demonstration. A quantitative evaluation of their outputs showed that both systems were capable of meeting surface anomaly detection requirements. The CCTV system was superior in its ability to collect, suppress and present data into a more useful form for the experimenters

Alternative Practices to Improve Surface Fleet Fuel Efficiency

Science.gov (United States)

2014-09-01

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

BUC implementation in Slovakia

International Nuclear Information System (INIS)

Chrapciak, V.; Vaclav, J.

2009-01-01

Improved calculation methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate criticality safety margin. Application of burnup credit (BUC) requires knowledge of the reactivity state of the irradiated fuel for which BUC is taken. The isotopic inventory and reactivity has to be calculated with validated codes. We use in Slovakia Gd 2 fuel with maximal enrichment of fuel pins 4.4%. Our transport and storage basket KZ-48 with boron steel is licensed for fresh fuel with enrichment 4.4%. In near future (2011 or 2012) we will use a new fuel with maximal enrichment of fuel pins 4.9%. For this fuel we plan to use existing KZ-48 with BUC application. In cooperation with Slovak Nuclear Regulatory Authority (UJD) we have started several years ago process of BUC implementation in Slovakia for VVER-440 reactors. We have already prepared methodology according IAEA methodology. We have validated computational systems (SCALE 5.1 already, SCALE 6 in progress). UJD will prepare regulation about BUC application in Slovakia. Last item is preparation of safety reports (for transport and storage) for the new fuel with average enrichment 4.87% in basket KZ-48 with BUC application.

HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

International Nuclear Information System (INIS)

Margeanu, C.A.; Margeanu, S.; Barbos, D.

2009-01-01

INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

Science.gov (United States)

Feddema, Rick

Feddema, Rick T. M.S.M.E., Purdue University, December 2013. Effect of Aviation Fuel Type and Fuel Injection Conditions on the Spray Characteristics of Pressure Swirl and Hybrid Air Blast Fuel Injectors. Major Professor: Dr. Paul E. Sojka, School of Mechanical Engineering Spray performance of pressure swirl and hybrid air blast fuel injectors are central to combustion stability, combustor heat management, and pollutant formation in aviation gas turbine engines. Next generation aviation gas turbine engines will optimize spray atomization characteristics of the fuel injector in order to achieve engine efficiency and emissions requirements. Fuel injector spray atomization performance is affected by the type of fuel injector, fuel liquid properties, fuel injection pressure, fuel injection temperature, and ambient pressure. Performance of pressure swirl atomizer and hybrid air blast nozzle type fuel injectors are compared in this study. Aviation jet fuels, JP-8, Jet A, JP-5, and JP-10 and their effect on fuel injector performance is investigated. Fuel injector set conditions involving fuel injector pressure, fuel temperature and ambient pressure are varied in order to compare each fuel type. One objective of this thesis is to contribute spray patternation measurements to the body of existing drop size data in the literature. Fuel droplet size tends to increase with decreasing fuel injection pressure, decreasing fuel injection temperature and increasing ambient injection pressure. The differences between fuel types at particular set conditions occur due to differences in liquid properties between fuels. Liquid viscosity and surface tension are identified to be fuel-specific properties that affect the drop size of the fuel. An open aspect of current research that this paper addresses is how much the type of aviation jet fuel affects spray atomization characteristics. Conventional aviation fuel specifications are becoming more important with new interest in alternative

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

International Nuclear Information System (INIS)

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

2013-01-01

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Impregnation/Agglomeration Laboratory Tests of Heavy Fuel from Prestige to Improve Its Manageability and Removal from Seawater Surface. (Physical Behaviour of Fuel Agglomates); Ensayos a Nivel de Laboratorio de Impregnacion/Aglomeracion del Fuel Procedente del Prestige para Facilitar su Manipulacion y Recogida en la Superficie del Mar (Comportamiento Fisico de los Aglomerados de Fuel)

Energy Technology Data Exchange (ETDEWEB)

Garcia Frutos, F. J.; Rodriguez, V.; Otero, J.

2002-07-01

The handling and removal problems showed by heavy fuel floating in seawater could be improved or solved by using materials that agglomerate it. These materials must fulfill the following condition: be inert materials in marine environment, the agglomerated fuel/material should float and its application and removal should be done using simple technologies. Based on these requirements, clay minerals, pine chips, mineral coal and charcoal were selected. The preliminary results on impregnation/agglomeration with the materials mentioned above of heavy fuel from Prestige at lab scale are presented in this paper. The results have shown that only hydrophobic materials, such as mineral coal and charcoal, are able to agglomerate with fuel, which is also a hydrophobic substance. Whereas the agglomerates fuel/mineral coal sink, the agglomerates fuel/charcoal keep floating on water surface. It can be concluded that the addition of charcoal on dispersed fuel in seawater could improve its handling and removal. In this sense, pilot scale and eventually controlled in situ tests to study the feasibility of the proposed solution should be performed. (Author) 2 refs.

Improving the corrosion resistance of proton exchange membrane fuel cell carbon supports by pentafluorophenyl surface functionalization

Science.gov (United States)

Forouzandeh, Farisa; Li, Xiaoan; Banham, Dustin W.; Feng, Fangxia; Joseph Kakanat, Abraham; Ye, Siyu; Birss, Viola

2018-02-01

In this study, the effect of surface functionalization on the electrochemical corrosion resistance of a high surface area, mesoporous colloid imprinted carbon powder (CIC), as well as microporous Vulcan carbon (VC, serving as the benchmark), was demonstrated, primarily for PEM fuel cell applications. CIC-22, which is highly hydrophilic and was synthesized with 22 nm silica colloid templates, and as-received, mildly hydrophobic, VC powders, were functionalized with 2,3,4,5,6-pentafluorophenyl (-PhF5) surface groups using a straightforward diazonium reduction reaction. These carbons were then subjected to corrosion testing, involving a potential cycling-step sequence in room temperature 0.5 M H2SO4. Using cyclic voltammetry and charge/time analysis, the double layer and pseudo-capacitive gravimetric charges of the carbons, prior to and after the application of these potential steps, were tracked in order to obtain information about surface area changes and the extent of carbon oxidation, respectively. It is shown that the corrosion resistance was improved by ca. 50-80% by surface functionalization, likely due to a combination of surface passivation (loss of carbon active sites) and increased surface hydrophobicity.

Estimates of Particulate Mass in Multi-Canister Overpacks

International Nuclear Information System (INIS)

SLOUGHTER, J.P.

2000-01-01

High, best estimate, and low values are developed for particulate inventories within MCO baskets that have been loaded with freshly cleaned fuel assemblies and scrap. These per-basket estimates are then applied to all anticipated MCO payload configurations to identify which configurations are bounding for each type of particulate. Finally the resulting bounding and nominal values for residual particulates are combined with corresponding values [from other documents] for particulates that may be generated by corrosion of exposed uranium after the fuel has been cleaned. The resulting rounded nominal estimate for a typical MCO after 40 years of storage is 8 kg. The estimate for a bounding total particulate case MCO is that it may contain up to 64 kg of particulate after 40 years of storage

Estimates of particulate mass in multi-canister overpacks

International Nuclear Information System (INIS)

SLOUGHTER, J.P.

1999-01-01

High, best estimate, and low values are developed for particulate inventories within MCO baskets that have been loaded with freshly cleaned fuel assemblies and scrap. These per-basket estimates are then applied to all anticipated MCO payload configurations to identify which configurations are bounding for each type of particulate. Finally the resulting bounding and nominal values for residual particulates are combined with corresponding values [from other documents] for particulate that may be generated by corrosion of exposed uranium after the fuel has been cleaned. The resulting rounded nominal estimate for a typical MCO after 40 years of storage is 8 kg. The estimate for a bounding total particulate case MCO is that it may contain up to 64 kg of particulate after 40 years of storage

Estimates of particulate mass in multi-canister overpacks

Energy Technology Data Exchange (ETDEWEB)

SLOUGHTER, J.P.

1999-02-25

High, best estimate, and low values are developed for particulate inventories within MCO baskets that have been loaded with freshly cleaned fuel assemblies and scrap. These per-basket estimates are then applied to all anticipated MCO payload configurations to identify which configurations are bounding for each type of particulate. Finally the resulting bounding and nominal values for residual particulates are combined with corresponding values [from other documents] for particulate that may be generated by corrosion of exposed uranium after the fuel has been cleaned. The resulting rounded nominal estimate for a typical MCO after 40 years of storage is 8 kg. The estimate for a bounding total particulate case MCO is that it may contain up to 64 kg of particulate after 40 years of storage.

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

Science.gov (United States)

1984-10-01

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

Use of a commercial heat transfer code to predict horizontally oriented spent fuel rod surface temperatures

International Nuclear Information System (INIS)

Wix, S.D.; Koski, J.A.

1993-03-01

Radioactive spent fuel assemblies are a source of hazardous waste that will have to be dealt with in the near future. It is anticipated that the spent fuel assemblies will be transported to disposal sites in spent fuel transportation casks. In order to design a reliable and safe transportation cask, the maximum cladding temperature of the spent fuel rod arrays must be calculated. A comparison between numerical calculations using commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degree C and 23 degree C for the low heat dissipation and high heat dissipation, respectively. The temperature predictions using helium as a fill gas are lower for the low and medium heat dissipation levels, but higher at the high heat dissipation. The temperature differences are 1 degree C and 6 degree C for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degree C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include experimental uncertainty in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This work demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects will be increasingly important as the amount of dissipated heat increases

Optimization of Reactor Temperature and Catalyst Weight for Plastic Cracking to Fuels Using Response Surface Methodology

Directory of Open Access Journals (Sweden)

Istadi Istadi

2011-01-01

Full Text Available The present study deals with effect of reactor temperature and catalyst weight on performance of plastic waste cracking to fuels over modified catalyst waste as well as their optimization. From optimization study, the most operating parameters affected the performance of the catalytic cracking process is reactor temperature followed by catalyst weight. Increasing the reactor temperature improves significantly the cracking performance due to the increasing catalyst activity. The optimal operating conditions of reactor temperature about 550 oC and catalyst weight about 1.25 gram were produced with respect to maximum liquid fuel product yield of 29.67 %. The liquid fuel product consists of gasoline range hydrocarbons (C4-C13 with favorable heating value (44,768 kJ/kg. ©2010 BCREC UNDIP. All rights reserved(Received: 10th July 2010, Revised: 18th September 2010, Accepted: 19th September 2010[How to Cite: I. Istadi, S. Suherman, L. Buchori. (2010. Optimization of Reactor Temperature and Catalyst Weight for Plastic Cracking to Fuels Using Response Surface Methodology. Bulletin of Chemical Reaction Engineering and Catalysis, 5(2: 103-111. doi:10.9767/bcrec.5.2.797.103-111][DOI: http://dx.doi.org/10.9767/bcrec.5.2.797.103-111 || or local:  http://ejournal.undip.ac.id/index.php/bcrec/article/view/797

Nuclear fuel element

International Nuclear Information System (INIS)

1974-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Nuclear fuel element

International Nuclear Information System (INIS)

Thompson, J.R.; Rowland, T.C.

1976-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Aviation investigation report : hard landing : fuel leak and fire[Sundance Balloons International Firefly 12B (hot air balloon) C-FNVM, Winnipeg, Manitoba, 15 nm NE, 11 August 2007

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-08-15

This investigation report discussed an incident in Winnipeg in which a hot air balloon attempting to land during strong winds was dragged on its side for approximately 700 feet. The balloon's burners struck the ground as the balloon came to a stop, after which a propane leak occurred. An intense, uncontrolled fire ensued as balloon passengers were exiting from the partially-inverted basket. The pilot and 2 passengers suffered serious injuries, while another 4 passengers suffered minor injuries. The balloon's 2 propane tanks and a fire extinguisher canister exploded during the fire, which destroyed the balloon's basket. The rigging of the balloon was examined and no failures were discovered. Pressure tests showed that the balloon's hoses started leaking at the crimped sleeve fittings at 150 psi. While the pilot had been informed of potentially heavy winds and thunderstorms, changes in wind speed and direction occurred earlier than the forecasted time of 10:00. It was concluded that standards are needed to ensure balloon cabin safety. Balloon operators do not currently require the use of protective helmets or gloves in case of dragged landings. A review will be conducted to address the issue of proposed emergency fuel shut-offs for balloons carrying fare-paying passengers. 2 figs.

Fuel Pellets from Wheat Straw: The Effect of Lignin Glass Transition and Surface Waxes on Pelletizing Properties

Science.gov (United States)

Wolfgang Stelte; Craig Clemons; Jens K. Holm; Jesper Ahrenfeldt; Ulrik B. Henriksen; Anand R. Sanadi

2012-01-01

The utilization of wheat straw as a renewable energy resource is limited due to its low bulk density. Pelletizing wheat straw into fuel pellets of high density increases its handling properties but is more challenging compared to pelletizing wood biomass. Straw has a lower lignin content and a high concentration of hydrophobic waxes on its outer surface that may limit...

Dependence of the specific surface area of the nuclear fuel with the matrix oxidation

International Nuclear Information System (INIS)

Gomez, F.; Quinones, J.; Iglesias, E.; Rodriguez, N.

2008-01-01

This paper is focused on the study of the changes in the specific surface area measured using BET techniques. The objective is to obtain a relation between this parameter and the change in the matrix stoichiometry (i.e., oxidation increase). None of the actual models used for extrapolating the behaviour of the spent fuel matrix under repository conditions have included this dependence yet. In this work the specific surface area of different uranium oxide were measured using N 2 (g) and Kr(g). The starting material was UO 2+x (s) with a size powder distribution lower than 20 μm. The results included in this paper shown a strong dependence on specific surface area with the matrix stoichiometry, i.e., and increase of more than one order of magnitude (SUO 2 = 6 m 2 *g -1 and SU 3 O 8 = 16.07 m 2 *g -1 ). Furthermore, the particle size distribution measured as a function of the thermal treatment done shows changes on the powder size related to the changes observed in the uranium oxide stoichiometry. (authors)

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

Spacers for use in nuclear fuel assembly

International Nuclear Information System (INIS)

Shiohata, Hironori; Nakamura, Shozo; Hasegawa, Kunio; Higuchi, Shigeo; Nagashima, Hideaki; Kawada, Yoshishige.

1987-01-01

Purpose: To prevent liquid film breakage at the surface of a fuel rod due to swirlings of steam flow generated at the upstream of a contact portion between the fuel rod and a spacer leaf spring, that is, below the contact portion. Constitution: Steam-hot water 2-phase streams flowing from the lower to the upper portions of a fuel assembly is hindered by leaf springs, thereby forming swirlings in the steam flow at the upstream of a contact portion between the fuel rod and the leaf springs, that is, below the contact portion. The horseshoe-like swirlings shed the liquid films at the surface of the fuel rod to remarkably decrease the heat cooling performance, by which the surface temperature of a fuel can is temporarily increased thereby possibly causing failures due to so-called burnout in view of the above, steps are formed to the spacer leaf spring for use in the fuel assembly, to reduce the pressure difference between the leaf spring and the fuel rod at the upstream of the springs relative to the 2-phase coolant stream. In this way, formation of the swirlings is moderated to prevent the liquid film breakage and improve the critical heat power. (Kamimura, M.)

Scoping corrosion tests on candidate waste package basket materials for the Yucca Mountain project

International Nuclear Information System (INIS)

Konynenburg, R.A. van; Curtis, P.G.; Summers, T.S.E.

1998-03-01

A scoping corrosion test was performed on candidate waste package basket materials. The corrosion medium was a pH-buffered solution of chemical species expected to be produced by radiolysis. The test was conducted at 90 C for 96 hours. Samples included aluminum-, copper-, stainless steel- and zirconium-based metallic materials and several ceramics, incorporating neutron-absorbing elements. Sample weight losses and solution chemical changes were measured. Both corrosion of the host materials and dissolution of the neutron-absorbing elements were studied. The ceramics and the zirconium-based materials underwent only minor corrosion. The stainless steel-based materials performed well except for a welded sample. The aluminum- and copper-based materials exhibited the highest corrosion rates. Boron dissolution depends on its chemical form. Boron oxide and many metal borides dissolve readily in acidic solutions while high-chromium borides and boron carbide, though thermodynamically unstable, exhibit little dissolution in short times. The results of solution chemical analyses were consistent with this. Gadolinium did not dissolve significantly from monazite, and hafnium showed little dissolution from a variety of host materials, in keeping with its low solubility

Spent fuel storage rack for BWR fuel assemblies

International Nuclear Information System (INIS)

Machado, O.; Henry, C.W.; Congleton, R.L.; Flynn, W.M.

1990-01-01

This patent describes for the use in storing nuclear fuel assemblies in a storage pool containing a coolant and having a pool floor, a fuel rack module. It comprises: a base plate to be disposed generally horizontally on the floor and having a horizontal surface area sufficient to support a fuel assemblies; uniformly spaced openings in the base plate, disposed in rows and columns throughout the surface area; fabricated cells of rectangular cross section extending over alternate openings along each row of the openings, the fabricated cells of each row being uniformly staggered by one opening with respect to the cells of its just adjacent rows so that the fabricated cells form a checkerboard like array; each of the fabricated cells having elongated walls mounted generally vertically on the base plate; each of the corners formed by the walls of each fabricated cell, which corners are internal of the periphery of the array, being disposed as closely adjacent as practicable to and face-to-face with a corner of an adjacent fabricated cell and joined by weld means so that substantially no space exists between adjacent cells. The cells being welded to their bottom ends to the base plate so that a strong compact modular structure is produced; neutron-absorbing means on the external surface of the fabricated cell walls except on the coextensive sections of the outer wall around the periphery of the array; and leveling pads are mounted under the base plate near the periphery thereof and adjustably engage the pool floor and intermediate leveling pads are mounted under cells within the fuel-rack module, the intermediate pads being uniformly disposed

New camera systems for fuel services

International Nuclear Information System (INIS)

Hummel, W.; Beck, H.J.

2010-01-01

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

Surface-to-surface biofilm transfer: a quick and reliable startup strategy for mixed culture microbial fuel cells.

Science.gov (United States)

Vogl, Andreas; Bischof, Franz; Wichern, Marc

2016-01-01

The startup of microbial fuel cells (MFCs) is known to be prone to failure or result in erratic performance impeding the research. The aim of this study was to advise a quick launch strategy for laboratory-scale MFCs that ensures steady operation performance in a short period of time. Different startup strategies were investigated and compared with membraneless single chamber MFCs. A direct surface-to-surface biofilm transfer (BFT) in an operating MFC proved to be the most efficient method. It provided steady power densities of 163 ± 13 mWm(-2) 4 days after inoculation compared to 58 ± 15 mWm(-2) after 30 days following a conventional inoculation approach. The in situ BFT eliminates the need for microbial acclimation during startup and reduces performance fluctuations caused by shifts in microbial biodiversity. Anaerobic pretreatment of the substrate and addition of suspended enzymes from an operating MFC into the new MFC proved to have a beneficial effect on startup and subsequent operation. Polarization methods were applied to characterize the startup phase and the steady state operation in terms of power densities, internal resistance and power overshoot during biofilm maturation. Applying this method a well-working MFC can be multiplied into an array of identically performing MFCs.

Thoracoamniotic shunting for fetal pleural effusion with hydropic change using a double-basket catheter: An insight into the preoperative determinants of shunting efficacy.

Science.gov (United States)

Hidaka, Nobuhiro; Kido, Saki; Sato, Yuka; Murata, Masaharu; Fujita, Yasuyuki; Kato, Kiyoko

2018-02-01

Although the efficacy of thoracoamniotic shunting (TAS) for fetal hydrothorax is well-recognized, the coexistence of hydrops fetalis is still a clinical challenge. The preoperative determinants of shunting efficacy are not fully understood. In this study, we aimed to investigate the perinatal and postnatal outcomes of hydrops fetalis with pleural effusion treated by TAS using a double-basket catheter, and to discuss the preoperative factors predictive of patients who will benefit from TAS. We conducted a retrospective study in hydropic fetuses with pleural effusion treated by TAS between 2007 and 2015. We extracted information regarding postnatal survival and pretherapeutic sonographic findings, including skin-edema thickness, pleural-effusion pocket size, and Doppler readings. Twelve subjects underwent TAS at a median gestational age of 29 +5 weeks (range, 25 +5 -33 +2 weeks). Skin edema disappeared or regressed in 7. Three experienced early neonatal death and the other 9 ultimately survived after a live birth at a median gestational age of 33 +4 weeks (range, 29 +1 -38 +2 weeks). All surviving children, except for 1, had a pretherapeutic pleural-effusion pocket greater than the precordial-edema thickness. All 3 children that died had precordial-edema thickness equal to or greater than the size of the pleural-effusion pocket. We achieved a high survival rate (75%) using the double-basket technique. A greater pretherapeutic width of skin edema compared with the pleural-effusion pocket is possibly suggestive of a treatment-resistant condition and subsequent poor postnatal outcome. Copyright © 2017 Elsevier B.V. All rights reserved.

Criticality evaluation of BWR MOX fuel transport packages using average Pu content

International Nuclear Information System (INIS)

Mattera, C.; Martinotti, B.

2004-01-01

Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by a homogeneous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, COGEMA LOGISTICS has studied a new calculation method based on the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in our approach. With this new method, for the same package reactivity, the Pu-content allowed in the package design approval can be higher. The COGEMA LOGISTICS' new method allows, at the design stage, to optimise the basket, materials or geometry for higher payload, keeping the same reactivity

EpiBasket: how e-commerce tools can improve epidemiological preparedness

Science.gov (United States)

Xing, Weijia; Hejblum, Gilles; Valleron, Alain-Jacques

2013-01-01

Background Should an emerging infectious disease outbreak or an environmental disaster occur, the collection of epidemiological data must start as soon as possible after the event's onset. Questionnaires are usually built de novo for each event, resulting in substantially delayed epidemiological responses that are detrimental to the understanding and control of the event considered. Moreover, the public health and/or academic institution databases constructed with responses to different questionnaires are usually difficult to merge, impairing necessary collaborations. We aimed to show that e-commerce concepts and software tools can be readily adapted to enable rapid collection of data after an infectious disease outbreak or environmental disaster. Here, the ‘customers’ are the epidemiologists, who fill their shopping ‘baskets’ with standardised questions. Methods For each epidemiological field, a catalogue of questions is constituted by identifying the relevant variables based on a review of the published literature on similar circumstances. Each question is tagged with information on its source papers. Epidemiologists can then tailor their own questionnaires by choosing appropriate questions from this catalogue. The software immediately provides them with ready-to-use forms and online questionnaires. All databases constituted by the different EpiBasket users are interoperable, because the corresponding questionnaires are derived from the same corpus of questions. Results A proof-of-concept prototype was developed for Knowledge, Attitudes and Practice (KAP) surveys, which is one of the fields of the epidemiological investigation frequently explored during, or after, an outbreak or environmental disaster. The catalogue of questions was initiated from a review of the KAP studies conducted during or after the 2003 severe acute respiratory syndrome epidemic. Conclusion Rapid collection of standardised data after an outbreak or environmental disaster can be

Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

International Nuclear Information System (INIS)

Dobrevski, I.; Zaharieva, N.

2008-01-01

Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

Multi Canister Overpack (MCO) Design Report [SEC 1 Thru 3

Energy Technology Data Exchange (ETDEWEB)

GOLDMANN, L.H.

2000-02-29

The MCO is designed to facilitate the removal, processing and storage of the spent nuclear fuel currently stored in the East and West K-Basins. The MCO is a stainless steel canister approximately 24 inches in diameter and 166 inches long with cover cap installed. The shell and the collar which is welded to the shell are fabricated from 304/304L dual certified stainless steel for the shell and F304/F304L dual certified for the collar. The shell has a nominal thickness of 1/2 inch. The top closure consists of a shield plug with four processing ports and a locking ring with jacking bolts to pre-load a metal seal under the shield plug. The fuel is placed in one of four types of baskets, excluding the SPR fuel baskets, in the fuel retention basin. Each basket is then loaded into the MCO which is inside the transfer cask. Once all of the baskets are loaded into the MCO, the shield plug with a process tube is placed into the open end of the MCO. This shield plug provides shielding for workers when the transfer cask, containing the MCO, is lifted from the pool. After being removed from the pool, the locking ring is installed and the jacking bolts are tightened to pre-load the metal main closure seal. The cask is then sealed and the MCO taken to the Cold Vacuum Drying (CVD) facility for bulk water removal and vacuum drying through the process ports. Covers for the process ports may be installed or removed as needed per operating procedures. The MCO is then transferred to the Canister Storage Building (CSB), in the closed transfer cask. At the CSB, the MCO is then removed from the cask and becomes one of two MCOs stacked in a storage tube. MCOs will have a cover cap welded over the shield plug providing a complete welded closure. A number of MCOs may be stored with just the mechanical seal to allow monitoring of the MCO pressure, temperature, and gas composition.

BUC implementation in Slovakia

International Nuclear Information System (INIS)

Chrapciak, V.; Vaclav, J.

2009-01-01

Improved calculation methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate criticality safety margin. Application of burnup credit requires knowledge of the reactivity state of the irradiated fuel for which application of burnup credit is taken. The isotopic inventory and reactivity has to be calculated with validated codes. We use in Slovakia Gd2 fuel with maximal enrichment of fuel pins 4.4%. Our transport and storage basket KZ-48 with boron steel is licensed for fresh fuel with enrichment 4.4%. In near future (2011 or 2012) we will use a new fuel with maximal enrichment of fuel pins 4.9%. For this fuel we plan to use existing KZ-48 with application of burnup credit application. In cooperation with Slovak Nuclear Regulatory Authority we have started several years ago process of application of burnup credit implementation in Slovakia for WWER-440 reactors. We have already prepared methodology according IAEA methodology. We have validated computational systems (SCALE 5.1 already, SCALE 6 in progress). Slovak Nuclear Regulatory Authority will prepare regulation about application of burnup credit application in Slovakia. Last item is preparation of safety reports (for transport and storage) for the new fuel with average enrichment 4.87% in basket KZ-48 with application of burnup credit application. (Authors)

Effects of salvage logging and pile-and-burn on fuel loading, potential fire behaviour, fuel consumption and emissions

Science.gov (United States)

Morris C. Johnson; Jessica E. Halofsky; David L. Peterson

2013-01-01

We used a combination of field measurements and simulation modelling to quantify the effects of salvage logging, and a combination of salvage logging and pile-and-burn fuel surface fuel treatment (treatment combination), on fuel loadings, fire behaviour, fuel consumption and pollutant emissions at three points in time: post-windstorm (before salvage logging), post-...

33 CFR 183.518 - Fuel tank openings.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Fuel tank openings. 183.518...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.518 Fuel tank openings. Each opening into the fuel tank must be at or above the topmost surface of the tank. ...

Nuclear fuel assembly

International Nuclear Information System (INIS)

Hirano, Yasushi; Hirukawa, Koji; Sakurada, Koichi.

1994-01-01

A bundle of fuel rods is divided into four fuel rod group regions of small fuel rod bundles by a cross-shaped partitioning structure consisting of paired plate-like structures which connect two opposing surfaces of a channel box. A water removing material with less neutron absorption (for example, Zr or a Zr alloy) or a solid moderator is inserted and secured to a portion of a non-boiling water region interposed between the paired plate-like structure. It has a structure that light water flows to the region in the plate-like structure. The volume, density or composition of the water removing material is controlled depending on the composition of the fuels, to change the moderating characteristics of neutrons in the non-boiling water region. This can easily moderate the difference of nuclear characteristics between each of fuel assemblies using fuel materials of different fuel compositions. Further, the reactivity control effect of the burnable poisons can be enhanced without worsening fuel economy or linear power density. (I.N.)

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

Surface Area Expansion of Electrodes with Grass-like Nanostructures to Enhance Electricity Generation in Microbial Fuel Cells

DEFF Research Database (Denmark)

Al Atraktchi, Fatima Al-Zahraa; Zhang, Yifeng; Noori, Jafar Safaa

2012-01-01

Microbial fuel cells (MFCs) have applications possibilities for wastewater treatment, biotransformation, and biosensor, but the development of highly efficient electrode materials is critical for enhancing the power generation. Two types of electrodes modified with nanoparticles or grass-like nan......Microbial fuel cells (MFCs) have applications possibilities for wastewater treatment, biotransformation, and biosensor, but the development of highly efficient electrode materials is critical for enhancing the power generation. Two types of electrodes modified with nanoparticles or grass...... of plain silicium showed a maximum power density of 86.0 mW/m2. Further expanding the surface area of carbon paper electrodes with gold nanoparticles resulted in a maximum stable power density of 346.9 mW/m2 which is 2.9 times higher than that achieved with conventional carbon paper. These results show...

Developing cask designs in the USSR

International Nuclear Information System (INIS)

Anon.

1989-01-01

To tackle the problem of transporting spent fuel from its VVER-1000s, the Soviet Union has developed two casks - the TK-10 and the TK-13 which are described here. Future developments of these designs may use a silicon-organics based material for the solid neutron shielding and no neutron absorbers in the fuel assembly basket. (author)

Experimental irradiation of UMo fuel: Pie results and modeling of fuel behaviour

International Nuclear Information System (INIS)

Languille, A.; Plancq, D.; Huet, F.; Guigon, B.; Lemoine, P.; Sacristan, P.; Hofman, G.; Snelgrove, J.; Rest, J.; Hayes, S.; Meyer, M.; Vacelet, H.; Leborgne, E.; Dassel, G.

2002-01-01

Seven full-sized U Mo plates containing ca. 8 g/cm 3 of uranium in the fuel meat have been irradiated since the beginning of the French U Mo development program. The first three of them with 20% 235 U enrichment were irradiated at maximum surfacic power under 150 W/cm 2 in the OSIRIS reactor up to 50% burn-up and are under examination. Their global behaviour is satisfactory: no failure and a low swelling. The other four plates were irradiated in the HFR Petten at maximum surfacic power between 150 and 250 W/cm 2 with two enrichments 20 and 35%. The experiment was stopped after two cycles due to a fuel failure. The post- irradiation examinations were completed in 2001 in Petten. Examinations showed a correct behaviour of 20% enriched plates and an abnormal behaviour of the two other plates (35%-enriched) with a clad failure on the plate 4. The fuel failure appears to result from a combination of factors that led to high corrosion cladding and high fuel meat temperatures. (author)

UO2 Fuel pellet impurities, pellet surface roughness and n(18O)/n(16O) ratios, applied to nuclear forensic science

International Nuclear Information System (INIS)

Pajo, L.

2001-01-01

In the last decade, law enforcement has faced the problem of illicit trafficking of nuclear materials. Nuclear forensic science is a new branch of science that enables the identification of seized nuclear material. The identification is not based on a fixed scheme, but further identification parameters are decided based on previous identification results. The analysis is carried out by using traditional analysis methods and applying modern measurement technology. The parameters are generally not unambiguous and not self-explanatory. In order to have a full picture about the origin of seized samples, several identification parameters should be used together and the measured data should be compared to corresponding data from known sources. A nuclear material database containing data from several fabrication plants is installed for the purpose. In this thesis the use of UO 2 fabrication plant specific parameters, fuel impurities, fuel pellet surface roughness and oxygen isotopic ratio in UO 2 were investigated for identification purposes in nuclear forensic science. The potential use of these parameters as 'fingerprints' is discussed for identification purposes of seized nuclear materials. Impurities of the fuel material vary slightly according to the fabrication method employed and a plant environment. Here the impurities of the seized UO 2 were used in order to have some clues about the origin of the fuel material by comparing a measured data to nuclear database information. More certainty in the identification was gained by surface roughness of the UO 2 fuel pellets, measured by mechanical surface profilometry. Categories in surface roughness between a different fuel element type and a producer were observed. For the time oxygen isotopic ratios were determined by Thermal Ionisation Mass Speckometry (TIMS). Thus a TIMS measurement method, using U 16 O + and U 18 0 + ions, was developed and optimised to achieve precise oxygen isotope ratio measurements for the

Management number identification method for nuclear fuel assembly

International Nuclear Information System (INIS)

Furuya, Nobuo; Mori, Kazuma.

1995-01-01

In the present invention, a management number indicated to appropriate portions of a fuel assembly can be read with no error for the management of nuclear fuel materials in the nuclear fuel assembly (counting management) and physical protection: PP. Namely, bar codes as a management number are printed by electrolytic polishing to one or more portions of a side surface of an upper nozzle of the assembly, an upper surface of a clamp and a side surface of a lower nozzle. The bar codes are read by a reader at one or more portions in a transporting path for transporting the fuel assembly and at a fuel detection device disposed in a fuel storage pool. The read signals are inputted to a computer. With such procedures, the nuclear fuel assembly can be identified with no error by reading the bar codes and without applying no danger to a human body. Since the reader is disposed in the course of the transportation and test for the assembly, and the read signals are inputted to the computer, the management for the counting number and PP is facilitated. (I.S.)

33 CFR 183.550 - Fuel tanks: Installation.

Science.gov (United States)

2010-07-01

...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Manufacturer Requirements § 183.550 Fuel tanks: Installation. (a) Each fuel tank must not be integral with any boat structure or mounted on an engine. (b) Each... the top surface of each metallic fuel tank when the boat is in its static floating position. (e) Each...

Fuel number identification method and device therefor

International Nuclear Information System (INIS)

Doi, Takami; Seno, Makoto; Tanaka, Keiji

1998-01-01

The present invention provides a method of and a device for automatically identifying the number on the upper surface of a fuel of a fuel assembly in a PWR type reactor. Namely, the number on the upper surface of the fuel assembly of the PWR is not arranged in a row, but indent letters are dispersed to predetermined positions of the surface to be indented. Accordingly, the identification of letters is difficult. In the present invention, the letters are identified by the following procedures. Procedure (1): the letters are detected while having a corner portion of the upper surface of a fuel assembly where the number is indented as characteristic points. A procedure (2): a letter region is determined to a relative position based on the characteristic points while determining indent letters having the same direction as one group. A procedure (3): a letter identification treatment is applied to the letter images in the above-mentioned letter region to identify them. A neural network is used for the letter identification treatment. (N.H.)

Fuel cells fuelled by Saccharides

International Nuclear Information System (INIS)

Schechner, P.; Mor, L.; Sabag, N.; Rubin, Z.; Bubis, E.

2005-01-01

Full Text:Saccharides, like glucose, fructose and lactose, are ideal renewable fuels. They have high energy content, are safe, transportable, easy to store, non-flammable, non poisonous, non-volatile, odorless, easy to produce anywhere and abundant. Fuel Cells are electro-chemical devices capable to convert chemical energy into electrical energy from fuels, with theoretical efficiencies higher than 0.8 at room temperatures and with low pollutant emissions. Fuel Cells that can produce electricity form saccharides will be able to replace batteries, power electrical plants from biomass wastes, and serve as engines for transportation. In spite of these advantages, saccharide fuelled fuel cells are no available yet. Two obstacles hinder the feasibility of this potentially revolutionary device. The first is the high stability of the saccharides, which requires a good catalyst to extract the electrons from the saccharide fuel. The second is related to the nature of the Fuel Cells: the physical process takes place at the interface surface between the fuel and the electrode. In order to obtain high densities, materials with high surface to volume ratio are needed. Efforts to overcome these obstacles will be described. The use of saccharides as a fuel was treated from the thermodynamic point of view and compared with other common fuels currently used in fuel cells. We summarize measurements performed in a membrane less Alkaline Fuel Cell, using glucose as a fuel and KOH as electrolyte. The anode has incorporated platinum particles and operated at room temperature. Measurements were done, at different concentrations of glucose, of the Open Circuit Voltage, Polarization Curves and Power Density as function of the Current Density. The maximum Power Density reached was 0.61 mW/cm 2 when the Current density was 2.13 mA/cm 2 and the measured Open Circuit Voltage was 0.771 V

Premixed direct injection nozzle for highly reactive fuels

Science.gov (United States)

Ziminsky, Willy Steve; Johnson, Thomas Edward; Lacy, Benjamin Paul; York, William David; Uhm, Jong Ho; Zuo, Baifang

2013-09-24

A fuel/air mixing tube for use in a fuel/air mixing tube bundle is provided. The fuel/air mixing tube includes an outer tube wall extending axially along a tube axis between an inlet end and an exit end, the outer tube wall having a thickness extending between an inner tube surface having a inner diameter and an outer tube surface having an outer tube diameter. The tube further includes at least one fuel injection hole having a fuel injection hole diameter extending through the outer tube wall, the fuel injection hole having an injection angle relative to the tube axis. The invention provides good fuel air mixing with low combustion generated NOx and low flow pressure loss translating to a high gas turbine efficiency, that is durable, and resistant to flame holding and flash back.

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

International Nuclear Information System (INIS)

P. Bernot

2001-01-01

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management and Operating Contractor (CRWMS M and O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% 235 U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

Energy Technology Data Exchange (ETDEWEB)

P. Bernot

2001-02-27

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited

Fuel element

International Nuclear Information System (INIS)

Armijo, J.S.

1976-01-01

A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

Monitoring for fuel sheath defects in three shipments of irradiated CANDU nuclear fuel

International Nuclear Information System (INIS)

Johnson, H.M.

1978-01-01

Analyses of radioactive gases within the Pegase shipping flask were performed at the outset and at the completion of three shipments of irradiated nuclear fuel from the Douglas Point Generating Station to Whiteshell Nuclear Research Establishment. No increases in the concentration of active gases, volatiles or particulates were observed. The activity of the WR-1 bay water rose only marginally due to the storage of the fuel. Other tests indicated that minimal surface contamination was present. These data established that defects in fuel element sheaths did not arise during the transport or the handling of this irradiated fuel. The observation has significance for the prospect of irradiated nuclear fuel transfer and handling in preparation for storage or disposal. (author)

Nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1980-01-01

A bimetallic spacer means is cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The bimetallic spacer means in one embodiment of the invention includes a space grid formed, at least principally, of zircaloy to the external surface of which are attached a plurality of stainless steel strips. In another embodiment the strips are attached to fuel pins. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. (author)

A market basket survey of As, Zn and Se in rice imports in Qatar: health implications.

Science.gov (United States)

Rowell, Candace; Kuiper, Nora; Al-Saad, Khalid; Nriagu, Jerome; Shomar, Basem

2014-08-01

Qatar is dependent on importation of rice, its staple dish, and is therefore susceptible to compromises of food quality in the global market. This market basket study assesses potential health risks of As exposure from rice consumption in Qatar and examines its contribution to the recommended nutritional intakes (RNI) for Zn and Se. Fifty-six rice types and 12 products sold in Qatar were analyzed by ICP/MS. Mean concentrations and ranges were 96.2±54.1μg/kg (9.76-258μg/kg) for As; 12.5±5.35mg/kg (2.79-29.9mg/kg) for Zn and 103±113μg/kg (Qatar is not a significant route of As exposure but can contribute up to 100% and 50% of the RNI for Se and Zn, respectively. Results indicate that children in Qatar may be at elevated risk of arsenic exposure from rice-based infant cereals but more data is needed to obtain a definitive assessment. Copyright © 2014 Elsevier Ltd. All rights reserved.

Method of manufacturing a graphite coated fuel can

International Nuclear Information System (INIS)

Saito, Koichi; Uchida, Shunsuke.

1984-01-01

Purpose: To improve the close bondability and homogeneity of a graphite coating formed at the inner surface of a fuel can. Method: A coating containing graphite dispersed in a volatile organic solvent is used and a graphite coating is formed to the inner surface of a fuel can by way of a plunger method. After applying graphite coating, an inert gas is caused to flow at a certain flow rate to the inside of the fuel can horizontally rotaged so that gassification and evaporation of the volatile organic solvent contained in the graphite coating may be promoted. Since drying of the graphite coating coated to the inner surface of the fuel can thus be controlled, a graphite coating with satisfactory close bondability and homogeneity can be formed. (Kawakami, Y.)

Process for the fabrication of a nuclear fuel

International Nuclear Information System (INIS)

Hirose, Yasuo.

1970-01-01

Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)

Integral analysis of cavity pressurization in a fuel rod during an ULOF driven TOP with inclusion of surface tension effects on froth gas bubbles and variable cavity conditions due to fuel melting and ejection

International Nuclear Information System (INIS)

Royl, P.

1984-02-01

The transient cavity pressurization in an ULOF driven TOP excursion has been analyzed for the SPX-1 reactor with an equation of state that allows to simulate the contribution of small froth gas bubbles to the pressure build-up in a fuel pin with inclusion of restraints from surface tension. Calculations were performed for various bubble parameters. Estimates are made for effective gas availabilities at fuel melting which can be used in a cavity model with an ideal gas equation to arrive at similar pressure transients

Development of alternative materials for BWR fuel springs

International Nuclear Information System (INIS)

Uruma, Y.; Osato, T.; Yamazaki, K.

2002-01-01

Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

Device for identifying fuel assembly

International Nuclear Information System (INIS)

Imai, Tetsuo; Miyazawa, Tatsuo.

1982-01-01

Purpose: To accurately identify a symbol printed on a hanging tool at the upper part of a fuel assembly. Constitution: Optical fibers are bundled to prepare a detector which is disposed at a predetermined position on a hanging tool. This position is set by a guide. Thus, the light emitted from an illumination lamp arrives at the bottom of a groove printed on the upper surface of the tool, and is divided into a weak light reflected upwardly and a strong light reflected on the surface lower than the groove. When these lights are received by the optical fibers, the fibers corresponding to the grooved position become dark, and the fibers corresponding to the ungrooved position become bright. Since the fuel assembly is identified by the dark and bright of the optical fibers as symbols, different machining can be performed every fuel assembly on the upper surface of the tool. (Yoshihara, H.)

Designing a miniaturised heated stage for in situ optical measurements of solid oxide fuel cell electrode surfaces, and probing the oxidation of solid oxide fuel cell anodes using in situ Raman spectroscopy

KAUST Repository

Brightman, E.; Maher, R.; Offer, G. J.; Duboviks, V.; Heck, C.; Cohen, L. F.; Brandon, N. P.

2012-01-01

A novel miniaturised heated stage for in operando optical measurements on solid oxide fuel cell electrode surfaces is described. The design combines the advantages of previously reported designs, namely, (i) fully controllable dual atmosphere operation enabling fuel cell pellets to be tested in operando with either electrode in any atmosphere being the focus of study, and (ii) combined electrochemical measurements with optical spectroscopy measurements with the potential for highly detailed study of electrochemical processes; with the following advances, (iii) integrated fitting for mounting on a mapping stage enabling 2-D spatial characterisation of the surface, (iv) a compact profile that is externally cooled, enabling operation on an existing microscope without the need for specialized lenses, (v) the ability to cool very rapidly, from 600 °C to 300 °C in less than 5 min without damaging the experimental apparatus, and (vi) the ability to accommodate a range of pellet sizes and thicknesses. © 2012 American Institute of Physics.

Overview of the FUTURIX-FTA Irradiation Experiment in the Phénix Reactor

Energy Technology Data Exchange (ETDEWEB)

Heather J.M. Chichester; Steve L. Hayes; Kenneth J. McClellan; Jean-Luc Paul; Marc Masson; Stewart L. Voit; Fabienne Delage

2015-09-01

The Advanced Fuels Campaign utilizes the Advanced Test Reactor (ATR) for most of its irradiation testing. Cadmium-shrouded baskets are used in ATR to modify the neutron spectrum to simulate a fast reactor environment for the fuel. FUTURIX-FTA is an irradiation experiment conducted in the Phenix fast reactor in France. Results from FUTURIX-FTA and irradiation tests in ATR using identical fuel compositions will be compared to identify and evaluate any differences in fuel behavior due to differences in the irradiation source.

Thirty years of transport package development for spent fuels

International Nuclear Information System (INIS)

Cory, A.R.

2005-01-01

By June 2005, when shipments of spent fuel for reprocessing from Germany are concluded, BNFL flask types will have been responsible for transporting more than 2000 tonnes of heavy metal in Europe in the form of spent fuel. Several thousand more tonnes of spent fuel have been transported by sea from Japan over the last thirty years. The design of spent fuel packages has not stood still for that time. In order to anticipate the changing needs of the nuclear power generation industry, advances have been made both in package design and analysis. Thirty years ago spent fuel burnup and initial enrichment were considerably lower, which was reflected in the different demands placed on the shielding design of packages, and in the design of the internal basket to separate the fuel assemblies. Technical development of both 'wet' (water-filled cavity) and 'dry' packages has progressed in parallel, and the relative merits and peculiarities of each type is explored. BNFL has considerable experience in the operation of both types, and is well placed to comment on practical and functional issues associated with both types. While there have been certain evolutionary changes affecting package design, there have also been more significant changes in the Design Safety Case. These have sometimes been necessary to meet changes in IAEA Regulations, or the challenges posed by the regulators themselves. In other cases advantage has been taken of improvements in analytical techniques to demonstrate increased margins of operational safety. Where possible these margins have also been increased by other means, such as taking advantage of commercial trends to reduce package thermal loads. A key factor over the last thirty years has been the increasing influence of the Regulating Authorities and the development of the IAEA Regulations. The various Competent Authorities now tend to have a higher proportion of technical experts, often recruited from the nuclear industry, and are thus more able to

Nanofluidic fuel cell

Science.gov (United States)

Lee, Jin Wook; Kjeang, Erik

2013-11-01

Fuel cells are gaining momentum as a critical component in the renewable energy mix for stationary, transportation, and portable power applications. State-of-the-art fuel cell technology benefits greatly from nanotechnology applied to nanostructured membranes, catalysts, and electrodes. However, the potential of utilizing nanofluidics for fuel cells has not yet been explored, despite the significant opportunity of harnessing rapid nanoscale reactant transport in close proximity to the reactive sites. In the present article, a nanofluidic fuel cell that utilizes fluid flow through nanoporous media is conceptualized and demonstrated for the first time. This transformative concept captures the advantages of recently developed membraneless and catalyst-free fuel cell architectures paired with the enhanced interfacial contact area enabled by nanofluidics. When compared to previously reported microfluidic fuel cells, the prototype nanofluidic fuel cell demonstrates increased surface area, reduced activation overpotential, superior kinetic characteristics, and moderately enhanced fuel cell performance in the high cell voltage regime with up to 14% higher power density. However, the expected mass transport benefits in the high current density regime were constrained by high ohmic cell resistance, which could likely be resolved through future optimization studies.

High Performance Fuel Desing for Next Generation Pressurized Water Reactors

International Nuclear Information System (INIS)

Mujid S. Kazimi; Pavel Hejzlar

2006-01-01

The use of internally and externally cooled annular fuel rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and economic assessment. The investigation was conducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperature. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasibility issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density

Opportunities in the certification of metal storage casks for spent fuel transport

International Nuclear Information System (INIS)

Teer, B.R.

1988-01-01

The current regulatory climate in the Transportation Certification Branch of the Nuclear Regulatory Commission is such that the applicant for a Certificate of Compliance must submit a well prepared, comprehensive Safety Analysis Report which meets all requirements of Regulatory Guide. All methods and procedures for structural, nuclear and thermal analyses must be qualified and benchmarked. The materials used for the containment boundary, impact limiters, neutron and gamma shields and internal structures must have documented and reproducible properties and, ideally, be referred in a recognized national standard. And, finally, in most cases, the analyses must be supplemented by model or full scale test results. The paper discusses some recent Transnuclear experiences with the NRC Transportation Certification Branch in attempting to get the TN-BRP and TN-REG casks certified for the transport of spent fuel. The authors have utilized non-standard materials for the baskets and a modified material for the containment. The impact limiters are not of a conventional design. Some of the analytical procedures are proprietary codes which are not used by others in the industry

Surface-Activated Amorphous Alloy Fuel Electrodes for Methanol Fuel Cell

OpenAIRE

Asahi, Kawashima; Koji, Hashimoto; The Research Institute for Iron, Steel and Other Metals; The Research Institute for Iron, Steel and Other Metals

1983-01-01

Amorphous alloy electrodes for electrochemical oxidation of methanol and its derivatives were obtained by the surface activation treatment consisting of electrodeposition of zinc on as-quenched amorphous alloy substrates, heating at 200-300℃ for 30 min, and subsequently leaching of zinc in an alkaline solution. The surface activation treatment provided a new method for the preparation of a large surface area on the amorphous alloys. The best result for oxidation of methanol, sodium formate an...

Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

International Nuclear Information System (INIS)

Inozemtsev, V.; Ganguly, C.

2005-01-01

Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)

Development and application of special instrumentation for materials accountancy and process control in spent fuel recycle plants

International Nuclear Information System (INIS)

Clark, P.A.; Gardner, N.; Merrill, N.H.; Whitehouse, K.R.

1996-01-01

Safe and optimum operations of spent fuel recycle plants rely on the availability of real time measurement systems at key points in the process. More than thirty types of special instrument systems have been developed and commissioned on the THORP reprocessing plant at Sellafield. These systems are compiled together with the associated information on measurement purpose, measurement technique and plant performance. A number of these measurement systems are of interest to support Safeguards arrangements on the plant. A more detailed overview of two such instrument systems respectively within the Head End and Product Finishing Stages of THORP is provided. The first of these is the Hulls Monitor, based on high resolution gamma spectrometry, as well as active and passive neutron measurements, of the basket of leached fuel cladding. This provides vital data for criticality assurance, nuclear material accountancy and inventory determination for ultimate disposal of the cladding waste. The second system is the Plutonium Inventory Monitoring System (PIMS) which employs passive neutron counting from a distributed array of neutron detectors within the Pu Finishing Line. This provides a near real time estimate of Pu inventories both during operations and at clean out of the Finishing Line. Both the Hulls Monitor and PIMS technologies are applicable to MOX Fuel recycle. Both systems enhance the control of fissile material in key areas of the recycle process which are of interest to the Safeguards authorities. (author)

Boiling performance and material robustness of modified surfaces with multi scale structures for fuel cladding development

Energy Technology Data Exchange (ETDEWEB)

Jo, HangJin; Kim, Jin Man [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Yeom, Hwasung [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States); Lee, Gi Cheol [Department of Mechanical Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Kiyofumi, Moriyama; Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Sridharan, Kumar; Corradini, Michael [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States)

2015-09-15

Highlights: • We improved boiling performance and material robustness using surface modification. • We combined micro/millimeter post structures and nanoparticles with heat treatments. • Compactly-arranged micrometer posts had improved boiling performance. • CHF increased significantly due to capillary pumping by the deposited NP layers. • Sintering procedure increased mechanical strength of the NP coating surface. - Abstract: By regulating the geometrical characteristics of multi-scale structures and by adopting heat treatment for protective layer of nanoparticles (NPs), we improved critical heat flux (CHF), boiling heat transfer (BHT), and mechanical robustness of the modified surface. We fabricated 1-mm and 100-μm post structures and deposited NPs on the structured surface as a nano-scale structured layer and protective layer at the same time, then evaluated the CHF and BHT and material robustness of the modified surfaces. On the structured surfaces without NPs, the surface with compactly-arranged micrometer posts had improved CHF (118%) and BHT (41%). On the surface with structures on which NPs had been deposited, CHF increased significantly (172%) due to capillary pumping by the deposited NP layers. The heat treatment improved robustness of coating layer in comparison to the one of before heat treatment. In particular, low-temperature sintering increased the hardness of the modified surface by 140%. The increased mechanical strength of the NP coating is attributed to reduction in coating porosity during sintering. The combination of micrometer posts structures and sintered NP coating can increase the safety, efficiency and reliability of advanced nuclear fuel cladding.

Boiling performance and material robustness of modified surfaces with multi scale structures for fuel cladding development

International Nuclear Information System (INIS)

Jo, HangJin; Kim, Jin Man; Yeom, Hwasung; Lee, Gi Cheol; Park, Hyun Sun; Kiyofumi, Moriyama; Kim, Moo Hwan; Sridharan, Kumar; Corradini, Michael

2015-01-01

Highlights: • We improved boiling performance and material robustness using surface modification. • We combined micro/millimeter post structures and nanoparticles with heat treatments. • Compactly-arranged micrometer posts had improved boiling performance. • CHF increased significantly due to capillary pumping by the deposited NP layers. • Sintering procedure increased mechanical strength of the NP coating surface. - Abstract: By regulating the geometrical characteristics of multi-scale structures and by adopting heat treatment for protective layer of nanoparticles (NPs), we improved critical heat flux (CHF), boiling heat transfer (BHT), and mechanical robustness of the modified surface. We fabricated 1-mm and 100-μm post structures and deposited NPs on the structured surface as a nano-scale structured layer and protective layer at the same time, then evaluated the CHF and BHT and material robustness of the modified surfaces. On the structured surfaces without NPs, the surface with compactly-arranged micrometer posts had improved CHF (118%) and BHT (41%). On the surface with structures on which NPs had been deposited, CHF increased significantly (172%) due to capillary pumping by the deposited NP layers. The heat treatment improved robustness of coating layer in comparison to the one of before heat treatment. In particular, low-temperature sintering increased the hardness of the modified surface by 140%. The increased mechanical strength of the NP coating is attributed to reduction in coating porosity during sintering. The combination of micrometer posts structures and sintered NP coating can increase the safety, efficiency and reliability of advanced nuclear fuel cladding

High Performance Fuel Desing for Next Generation Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Mujid S. Kazimi; Pavel Hejzlar

2006-01-31

The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

Reactor fuel charging equipment

International Nuclear Information System (INIS)

Wade, Elman.

1977-01-01

In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

Effect of fuel type and deposition surface temperature on the growth and structure of ash deposit collected during co-firing of coal with sewage-sludge, saw-dust and refuse derived fuel

Energy Technology Data Exchange (ETDEWEB)

Kupka, Tomasz; Zajac, Krzysztof; Weber, Roman [Clausthal Univ. of Technology, Clausthal-Zellerfeld (Germany). Inst. of Energy Process Engineering and Fuel Technology

2008-07-01

Blends of a South African bituminous ''Middleburg'' coal and three alternative fuels (a municipal sewage-sludge, a saw-dust and a refuse derived fuel) have been fired in the slagging reactor to examine the effect of the added fuel on slagging propensity of the mixtures. Two kinds of deposition probes have been used, un-cooled ceramic probes and air-cooled steal probes. Distinct differences in physical and chemical structures of the deposits collected using the un-cooled ceramic probes and air-cooled metal probes have been observed. Glassy, easily molten deposits collected on un-cooled ceramic deposition probes were characteristic for co-firing of municipal sewage-sludge with coal. Porous, sintered (not molten) but easily removable deposits of the same fuel blend have been collected on the air-cooled metal deposition probes. Loose, easy removable deposits have been sampled on air-cooled metal deposition probe during co-firing of coal/saw-dust blends. The mass of the deposit sampled at lower surface temperatures (550-700 C) was always larger than the mass sampled at higher temperatures (1100-1300 C) since the higher temperature ash agglomerated and sintered much faster than the low temperature deposit. (orig.)

Characteristics of fuel crud and its impact on storage, handling, and shipment of spent fuel

International Nuclear Information System (INIS)

Hazelton, R.F.

1987-09-01

Corrosion products, called ''crud,'' form on out-of-reactor surfaces of nuclear reactor systems and are transported by reactor coolant to the core, where they deposit on external fuel-rod cladding surfaces and are activated by nuclear reactions. After discharge of spent fuel from a reactor, spallation of radioactive crud from the fuel rods could impact wet or dry storage operations, handling (including rod consolidation), and shipping. It is the purpose of this report to review earlier (1970s) and more recent (1980s) literature relating to crud, its characteristics, and any impact it has had on actual operations. Crud characteristics vary from reactor type to reactor type, reactor to reactor, fuel assembly to fuel assembly in a reactor, circumferentially and axially in an assembly, and from cycle to cycle for a specific facility. To characterize crud of pressurized-water (PWRs) and boiling-water reactors (BWRs), published information was reviewed on appearance, chemical composition, areal density and thickness, structure, adhesive strength, particle size, and radioactivity. Information was also collected on experience with crud during spent fuel wet storage, rod consolidation, transportation, and dry storage. From experience with wet storage, rod consolidation, transportation, and dry storage, it appears crud spallation can be managed effectively, posing no significant radiological problems. 44 refs., 11 figs

High pressure combustion of liquid fuels. [alcohol and n-paraffin fuels

Science.gov (United States)

Canada, G. S.

1974-01-01

Measurements were made of the burning rates and liquid surface temperatures for a number of alcohol and n-paraffin fuels under natural and forced convection conditions. Porous spheres ranging in size from 0.64-1.9 cm O.D. were emloyed to simulate the fuel droplets. The natural convection cold gas tests considered the combustion in air of methanol, ethanol, propanol-1, n-pentane, n-heptane, and n-decane droplets at pressures up to 78 atmospheres. The pressure levels of the natural convection tests were high enough so that near critical combustion was observed for methanol and ethanol vaporization rates and liquid surface temperature measurements were made of droplets burning in a simulated combustion chamber environment. Ambient oxygen molar concentrations included 13%, 9.5% and pure evaporation. Fuels used in the forced convection atmospheric tests included those listed above for the natural convection tests. The ambient gas temperature ranged from 600 to 1500 K and the Reynolds number varied from 30 to 300. The high pressure forced convection tests employed ethanol and n-heptane as fuels over a pressure range of one to 40 atmospheres. The ambient gas temperature was 1145 K for the two combustion cases and 1255 K for the evaporation case.

Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

International Nuclear Information System (INIS)

Ellis, Ronald James

2011-01-01

permanent reflector. It is surrounded by a water reflector of effectively infinite thickness. In the axial direction, the reactor is reflected by water above and below the reactor. The irradiation facilities, one for UN and the other for UO 2 pellets, utilize a thin cylindrical hafnium shield approximately 4 cm in diameter surrounding the facility basket to reduce the thermal neutron flux sufficiently such that the linear power rating in the irradiated fuel pins will be similar to PWR operating conditions. The facilities each contain nine fuel pins, each comprising 10 fuel pellets, arranged as if three fuel rods.

Probable leaching mechanisms for spent fuel

International Nuclear Information System (INIS)

Wang, R.; Katayama, Y.B.

1981-01-01

At the Pacific Northwest Laboratory, researchers in the Waste/Rock Interaction Technology Program are studying spent fuel as a possible waste form for the Office of Nuclear Waste Isolation. This paper presents probable leaching mechanisms for spent fuel and discusses current progress in identifying and understanding the leaching process. During the past year, experiments were begun to study the complex leaching mechanism of spent fuel. The initial work in this investigation was done with UO 2 , which provided the most information possible on the behavior of the spent-fuel matrix without encountering the very high radiation levels associated with spent fuel. Both single-crystal and polycrystalline UO 2 samples were used for this study, and techniques applicable to remote experimentation in a hot cell are being developed. The effects of radiation are being studied in terms of radiolysis of water and surface activation of the UO 2 . Dissolution behavior and kinetics of UO 2 were also investigated by electrochemical measurement techniques. These data will be correlated with those acquired when spent fuel is tested in a hot cell. Oxidation effects represent a major area of concern in evaluating the stability of spent fuel. Dissolution of UO 2 is greatly increased in an oxidizing solution because the dissolution is then controlled by the formation of hexavalent uranium. In solutions containing very low oxygen levels (i.e., reducing solutions), oxidation-induced dissolution may be possible via a previously oxidized surface, through exposure to air during storage, or by local oxidants such as O 2 and H 2 O 2 produced from radiolysis of water and radiation-activated UO 2 surfaces. The effects of oxidation not only increase the dissolution rate, but could lead to the disintegration of spent fuel into fine fragments

High performance reliability fuel pellet

International Nuclear Information System (INIS)

Beuchel, P.H.; Lee, Y.C.

1989-01-01

A fuel pellet for a nuclear reactor fuel rod is described comprising: a substantially cylindrical central section; a convex first end section smoothly joined to one axial end of the central section at a first junction, the first junction approximating a smooth and continuous curved surface; a concave second end section joined to the central section at a second junction, the second junction approximating a smooth and continuous curved surface, wherein the curvature of the concave second end section is conformed to the curvature of the convex first end section

Possibility for dry storage of the WWR-K reactor spent fuel

International Nuclear Information System (INIS)

Arinkin, F.M.; Belyakova, E.A.; Gizatulin, Sh.Kh.; Khromushin, I.V.; Koltochik, S.N.; Maltseva, R.M.; Medvedeva, Z.V.; Petukhov, V.K.; Soloviev, Yu.A.; Zhotabaev, Zh.R.

2000-01-01

This work is devoted to development of the way for dry storage of spent fuel of the WWR-K reactor. Residual energy release in spent fuel element assembly was determined via fortune combination of calculations and experiments. The depth of fission product occurrence relative to the fuel element shroud surface was found experimentally. The time of fission product release to the fuel element shroud surface was estimated. (author)

Numerical modeling of laboratory-scale surface-to-crown fire transition

Science.gov (United States)

Castle, Drew Clayton

Understanding the conditions leading to the transition of fire spread from a surface fuel to an elevated (crown) fuel is critical to effective fire risk assessment and management. Surface fires that successfully transition to crown fires can be very difficult to suppress, potentially leading to damages in the natural and built environments. This is relevant to chaparral shrub lands which are common throughout parts of the Southwest U.S. and represent a significant part of the wildland urban interface. The ability of the Wildland-Urban Interface Fire Dynamic Simulator (WFDS) to model surface-to-crown fire transition was evaluated through comparison to laboratory experiments. The WFDS model is being developed by the U.S. Forest Service (USFS) and the National Institute of Standards and Technology. The experiments were conducted at the USFS Forest Fire Laboratory in Riverside, California. The experiments measured the ignition of chamise (Adenostoma fasciculatum) crown fuel held above a surface fire spreading through excelsior fuel. Cases with different crown fuel bulk densities, crown fuel base heights, and imposed wind speeds were considered. Cold-flow simulations yielded wind speed profiles that closely matched the experimental measurements. Next, fire simulations with only the surface fuel were conducted to verify the rate of spread while factors such as substrate properties were varied. Finally, simulations with both a surface fuel and a crown fuel were completed. Examination of specific surface fire characteristics (rate of spread, flame angle, etc.) and the corresponding experimental surface fire behavior provided a basis for comparison of the factors most responsible for transition from a surface fire to the raised fuel ignition. The rate of spread was determined by tracking the flame in the Smokeview animations using a tool developed for tracking an actual flame in a video. WFDS simulations produced results in both surface fire spread and raised fuel bed

Spent fuel and fuel pool component integrity. Annual report, FY 1979

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

1980-05-01

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project (SPAR-II)

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-01

conditions which would impact on the long-term spent fuel integrity. Vacuum drying and hot gas recirculation drying are employed in preparation leading to dry storage. The drying process is of importance to the dry storage systems as it prevents the presence of any water remnants which could potentially facilitate some degradation mechanisms during storage. Temperature increase during drying has to be controlled to minimize the potential for hydride re-orientation. In addition to research on spent fuel assembly integrity, research activities on the behaviours of storage facility components including reinforced concrete, neutron absorbers used in storage racks and baskets, neutron shielding materials, metal gaskets, polymer seal, stainless steel and carbon steel are also being conducted. As a conclusion, it remains important to continue to study and share results regarding fuel and materials behaviour in storage. In particular, changes in fuel and material design, and the increase in discharge burnup require continued research on spent fuel performance in long- term storage and post-storage transportation.

Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project (SPAR-II)

International Nuclear Information System (INIS)

2012-01-01

conditions which would impact on the long-term spent fuel integrity. Vacuum drying and hot gas recirculation drying are employed in preparation leading to dry storage. The drying process is of importance to the dry storage systems as it prevents the presence of any water remnants which could potentially facilitate some degradation mechanisms during storage. Temperature increase during drying has to be controlled to minimize the potential for hydride re-orientation. In addition to research on spent fuel assembly integrity, research activities on the behaviours of storage facility components including reinforced concrete, neutron absorbers used in storage racks and baskets, neutron shielding materials, metal gaskets, polymer seal, stainless steel and carbon steel are also being conducted. As a conclusion, it remains important to continue to study and share results regarding fuel and materials behaviour in storage. In particular, changes in fuel and material design, and the increase in discharge burnup require continued research on spent fuel performance in long- term storage and post-storage transportation.

Mechanical and temperature contact in fuel rod cladding

International Nuclear Information System (INIS)

Fredriksson, B.E.; Rydholm, S.G.

1977-01-01

The paper presents results for the effect of different types of slip rules on the contact stress distribution. It is shown that the contact shear stress is smaller for the hardening model than for the ideal model. It is also shown that a crack in the fuel increases the contact stresses and that at temperature decrease high tensile stresses arise after eventual welding. It is also shown how particles between fuel and cladding influence the stresses. Also here the effect of eventual welding is studied. The present method is well suited to study cracks and crack propagation. The surfaces of the existing cracks are defined as contact surfaces and the crack extension work is calculated by releasing the nodes at the crack tip. As the crack surfaces are defined as contact surfaces eventual crack closure is automatically taken into account. Crack extension work is calculated for existing cracks in the cladding. It is shown that cracks in the fuel and particles between fuel and cladding will increase the crack extension work

Method and device for taking out spent fuel

International Nuclear Information System (INIS)

Kono, Takayuki; Shimamoto, Takijiro; Otsuka, Shiomi.

1994-01-01

Both ends of a fuel assembly are cut and removed, and spacers are removed into a single body of a fuel cladding tube. A plurality of the fuel cladding tubes as single bodies are sent into a ring-like electrode. An arc forming electric power source is actuated, to form arcs between the fuel cladding tube and the ring-like electric power source. A magnetic field forming power source is actuated, to form magnetic fields around the ring-like electrode by solenoid coils. The direction of the magnetic fluxes in this magnetic fields intersect the arcs. Then, the arcs are applied with a rotational force by the magnetic fields formed by the solenoid coils and rotate on the surface of the fuel cladding tube and the inner surface of the ring-like electrode. The fuel cladding tube is exhausted circumferentially uniformly by the rotational arcs. Since spent fuels are taken out by melting and scattering the fuel cladding tube by the heat and force of the arcs, the operation is facilitated. (I.N.)

Nuclear fuel rods

International Nuclear Information System (INIS)

Wada, Toyoji.

1979-01-01

Purpose: To remove failures caused from combination of fuel-cladding interactions, hydrogen absorptions, stress corrosions or the likes by setting the quantity ratio of uranium or uranium and plutonium relative to oxygen to a specific range in fuel pellets and forming a specific size of a through hole at the center of the pellets. Constitution: In a fuel rods of a structure wherein fuel pellets prepared by compacting and sintering uranium dioxide, or oxide mixture consisting of oxides of plutonium and uranium are sealed with a zirconium metal can, the ratio of uranium or uranium and plutonium to oxygen is specified as 1 : 2.01 - 1 : 2.05 in the can and a passing hole of a size in the range of 15 - 30% of the outer diameter of the fuel pellet is formed at the center of the pellet. This increases the oxygen partial pressure in the fuel rod, oxidizes and forms a protection layer on the inner surface of the can to control the hydrogen absorption and stress corrosion. Locallized stress due to fuel cladding interaction (PCMI) can also be moderated. (Horiuchi, T.)

Nuclear fuel element

International Nuclear Information System (INIS)

Yamanaka, Tsuneyasu.

1976-01-01

Purpose: To provide a mechanism for the prevention of fuel pellet dislocation in fuel can throughout fuel fablication, fuel transportation and reactor operation. Constitution: A plenum spacer as a mechanism for the prevention of fuel pellet dislocation inserted into a cladding tube comprises split bodies bundled by a frame and an expansion body being capable of inserting into the central cavity of the split bodies. The expansion body is, for example, in a conical shape and the split bodies are formed so that they define in the center portion, when disposed along the inner wall of the cladding tube, a gap capable of inserting the conical body. The plenum spacer is assembled by initially inserting the split bodies in a closed state into the cladding tube after the loading of the pellets, pressing their peripheral portions and then inserting the expansion body into the space to urge the split bodies to the inner surface of the cladding tube. (Kawakami, Y.)

Fuel pin bowing and related investigation of WWER-440 control rod influence on power release inside of neighbouring fuel pins

International Nuclear Information System (INIS)

Mikus, J.

2005-01-01

The purpose of this work consists in investigation of the WWER-440 control rod (CR) influence on space power distribution, especially from viewpoint of the values and gradient occurrence that could result in static and cyclic loads with some consequences, e.g. fuel pin bowing. As known, CR can cause power peaks in periphery fuel pins of adjacent operating assemblies because of the butt joint design of the absorbing adapter to the CR fuel part, that is, presence of the water cavity resulting in a flash up of thermal neutrons. As a consequence, beside well-known peaks in axial power distribution, above power gradients can occur inside of mentioned fuel pins. Because of complicated geometry and material composition of the CR, the detailed calculations concerning both above phenomena are complicated, too. Therefore it is useful to acquire appropriate experimental data to investigate mentioned influence and compare them with calculations. Since detailed power distributions cannot be obtained in the NPP, needed information is provided by means of experiments on research reactors. In case of measurements inside of fuel pins, special (e.g. track) detectors placed between fuel pellets are used. Such works are relatively complicated and time consuming, therefore an evaluation based on mathematical modelling and numerical approximation was proposed by means of that, and using measured power release in some selected fuel pins, information about power release inside of one of these fuel pins, can be obtained. For this purpose, an experiment on light water, zero-power research reactor LR-0 was realized and axial power distribution measurements were performed in a WWER-440 type core near to an authentic CR model. Application of the above evaluation method is demonstrated on one ''investigated'' fuel pin neighbouring CR by means of following results: 1. Axial power distribution inside of investigated fuel pin in two opposite positions on its pellets surface that are situated to

In-situ study of the gas-phase composition and temperature of an intermediate-temperature solid oxide fuel cell anode surface fed by reformate natural gas

Science.gov (United States)

Santoni, F.; Silva Mosqueda, D. M.; Pumiglia, D.; Viceconti, E.; Conti, B.; Boigues Muñoz, C.; Bosio, B.; Ulgiati, S.; McPhail, S. J.

2017-12-01

An innovative experimental setup is used for in-depth and in-operando characterization of solid oxide fuel cell anodic processes. This work focuses on the heterogeneous reactions taking place on a 121 cm2 anode-supported cell (ASC) running with a H2, CH4, CO2, CO and steam gas mixture as a fuel, using an operating temperature of 923 K. The results have been obtained by analyzing the gas composition and temperature profiles along the anode surface in different conditions: open circuit voltage (OCV) and under two different current densities, 165 mA cm-2 and 330 mA cm-2, corresponding to 27% and 54% of fuel utilization, respectively. The gas composition and temperature analysis results are consistent, allowing to monitor the evolution of the principal chemical and electrochemical reactions along the anode surface. A possible competition between CO2 and H2O in methane internal reforming is shown under OCV condition and low current density values, leading to two different types of methane reforming: Steam Reforming and Dry Reforming. Under a current load of 40 A, the dominance of exothermic reactions leads to a more marked increase of temperature in the portion of the cell close to the inlet revealing that current density is not uniform along the anode surface.

Stability of yttria-stabilized zirconia during pyroprocessing tests

Energy Technology Data Exchange (ETDEWEB)

Choi, Eun-Young, E-mail: eychoi@kaeri.re.kr; Lee, Jeong; Lee, Sung-Jai; Kim, Sung-Wook; Jeon, Sang-Chae; Cho, Soo Haeng; Oh, Seung Chul; Jeon, Min Ku; Lee, Sang Kwon; Kang, Hyun Woo; Hur, Jin-Mok

2016-07-15

In this study, the feasibility of yttria-stabilized zirconia (YSZ) was investigated for use as a ceramic material, which can be commonly used for both electrolytic reduction and electrorefining. First, the stability of YSZ in salts for electrolytic reduction and electrorefining was examined. Then, its stability was demonstrated by a series of pyroprocessing tests, such as electrolytic reduction, LiCl distillation, electrorefining, and LiCl−KCl distillation, using a single stainless steel wire mesh basket containing fuel and YSZ. A single basket was used by its transportation from one test to subsequent tests without the requirements for unloading.

Nuclear fuel pellet loading machine

International Nuclear Information System (INIS)

Dazen, J.R.; Denero, J.V.

1976-01-01

A nuclear fuel pellet loading machine is described including an inclined rack mounted on a base and having parallel spaced grooves on its upper surface arranged to support fuel rods. A fuel pellet tray is adapted to be placed on a table spaced from the rack, the tray having columns of fuel pellets which are in alignment with the open ends of fuel rods located in the rack grooves. A transition plate is mounted between the fuel rod rack and the fuel pellet tray to receive and guide the pellets into the open ends of the fuel rods. The pellets are pushed into the fuel rods by a number of mechanical fingers mounted on a motor operated block which is moved along the pellet tray length by a drive screw driven by the motor. To facilitate movement of the pellets in the fuel rods the rack is mounted on a number of spaced vibrators which vibrate the fuel rods during fuel pellet insertion. A pellet sensing device movable into an end of each fuel rod indicates to an operator when each rod has been charged with the correct number of pellets

Research of fuel temperature control in fuel pipeline of diesel engine using positive temperature coefficient material

Directory of Open Access Journals (Sweden)

Xiaolu Li

2016-01-01

Full Text Available As fuel temperature increases, both its viscosity and surface tension decrease, and this is helpful to improve fuel atomization and then better combustion and emission performances of engine. Based on the self-regulated temperature property of positive temperature coefficient material, this article used a positive temperature coefficient material as electric heating element to heat diesel fuel in fuel pipeline of diesel engine. A kind of BaTiO3-based positive temperature coefficient material, with the Curie temperature of 230°C and rated voltage of 24 V, was developed, and its micrograph and element compositions were also analyzed. By the fuel pipeline wrapped in six positive temperature coefficient ceramics, its resistivity–temperature and heating characteristics were tested on a fuel pump bench. The experiments showed that in this installation, the surface temperature of six positive temperature coefficient ceramics rose to the equilibrium temperature only for 100 s at rated voltage. In rated power supply for six positive temperature coefficient ceramics, the temperature of injection fuel improved for 21°C–27°C within 100 s, and then could keep constant. Using positive temperature coefficient material to heat diesel in fuel pipeline of diesel engine, the injection mass per cycle had little change, approximately 0.3%/°C. This study provides a beneficial reference for improving atomization of high-viscosity liquids by employing positive temperature coefficient material without any control methods.

Nuclear fuel element end fitting

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

A typical embodiment of the invention has an array of sockets that are welded to the intersections of the plates that form the upper and lower end fittings of a nuclear reactor fuel element. The sockets, which are generally cylindrical in shape, are oriented in directions that enable the longitudinal axes of the sockets to align with the longitudinal axes of the fuel rods that are received in the respective sockets. Detents impressed in the surfaces of the sockets engage mating grooves that are formed in the ends of the fuel rods to provide for the structural integrity of the fuel element

Optimization of Biodiesel-Diesel Blended Fuel Properties and Engine Performance with Ether Additive Using Statistical Analysis and Response Surface Methods

Directory of Open Access Journals (Sweden)

Obed M. Ali

2015-12-01

Full Text Available In this study, the fuel properties and engine performance of blended palm biodiesel-diesel using diethyl ether as additive have been investigated. The properties of B30 blended palm biodiesel-diesel fuel were measured and analyzed statistically with the addition of 2%, 4%, 6% and 8% (by volume diethyl ether additive. The engine tests were conducted at increasing engine speeds from 1500 rpm to 3500 rpm and under constant load. Optimization of independent variables was performed using the desirability approach of the response surface methodology (RSM with the goal of minimizing emissions and maximizing performance parameters. The experiments were designed using a statistical tool known as design of experiments (DoE based on RSM.

Corrosion product deposition on fuel element surfaces of a boiling water reactor

International Nuclear Information System (INIS)

Orlov, A.

2011-01-01

solid solutions. These spinels are well-known for their magnetic behaviour. Since non magnetic zinc ferrite (ZnFe 2 O 4 ) may become magnetic when doped with even small amounts of Ni and/or Mn, their occurrence in the deposit layer has been analyzed. The magnetic permeability of zinc ferrite, trevorite (NiFe 2 O 4 ) and jacobsite (MnFe 2 O 4 ) and their solid solutions are estimated by magnetic moment additivity. The theoretical and literature data are compared with the magnetic permeability of the spinel-type fuel deposition layers gained by reactor pool side Eddy current (EC) analyses. The calculated thicknesses and magnetic permeability values of the deposition layers (estimated by MAGNACROX multifrequency EC method) are compared with the values estimated using an “ion magnetic moment additivity” model. The buildup of corrosion product deposits (crud) on the fuel cladding of the boiling water reactor, Kernkraftwerk Leibstadt (KKL) Switzerland has been investigated by using Gibbs Energy Minimization (GEM-Selector code) calculations of thermodynamic equilibrium at in situ temperatures and pressures, and by applying local experimental PIE techniques. Under the KKL water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)Fe 2 O 4 spinel solid solutions. GEM calculations applied to the boiling zone match with the EPMA findings, indicating that zinc-rich ferritic spinels are formed on KKL fuel cladding mainly at the lower pin elevations under Zn water chemistry conditions. For the prediction of the crud formation on the fuel cladding, Gibbs energy minimization of the element species during water evaporation in the BWR core were modelled. On the base of the literature data, the solid solution model is proposed and used for the modelling. The modelled processes of water evaporation propose that the spinel phase is formed only at large amount of the evaporated water locally at the pin surface. Finally, all the

Microfabrication of Microchannels for Fuel Cell Plates

Directory of Open Access Journals (Sweden)

Ho Su Jang

2009-12-01

Full Text Available Portable electronic devices such as notebook computers, PDAs, cellular phones, etc., are being widely used, and they increasingly need cheap, efficient, and lightweight power sources. Fuel cells have been proposed as possible power sources to address issues that involve energy production and the environment. In particular, a small type of fuel-cell system is known to be suitable for portable electronic devices. The development of micro fuel cell systems can be achieved by the application of microchannel technology. In this study, the conventional method of chemical etching and the mechanical machining method of micro end milling were used for the microfabrication of microchannel for fuel cell separators. The two methods were compared in terms of their performance in the fabrication with regards to dimensional errors, flatness, straightness, and surface roughness. Following microchannel fabrication, the powder blasting technique is introduced to improve the coating performance of the catalyst on the surface of the microchannel. Experimental results show that end milling can remarkably increase the fabrication performance and that surface treatment by powder blasting can improve the performance of catalyst coating.

Laser assisted decontamination of nuclear fuel elements

International Nuclear Information System (INIS)

Padma Nilaya, J.; Biswas, Dhruba J.; Kumar, Aniruddha

2010-04-01

Laser assisted removal of loosely bound fuel particulates from the clad surface following the process of pellet loading has decided advantages over conventional methods. It is a dry and noncontact process that generates very little secondary waste and can occur inside a glove box without any manual interference minimizing the possibility of exposure to personnel. The rapid rise of the substrate/ particulate temperature owing to the absorption of energy from the incident laser pulse results in a variety of processes that may lead to the expulsion of the particulates. As a precursor to the cleaning of the fuel elements, initial experiments were carried out on contamination simulated on commonly used clad surfaces to gain a first hand experience on the various laser parameters for which as efficient cleaning can be obtained without altering the properties of the clad surface. The cleaning of a dummy fuel element was subsequently achieved in the laboratory by integrating the laser with a work station that imparted simultaneous rotational and linear motion to the fuel element. (author)