Performance testing of thermal analysis codes for nuclear fuel casks

International Nuclear Information System (INIS)

Sanchez, L.C.

1987-01-01

In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

Experimental Study of Turbine Fuel Thermal Stability in an Aircraft Fuel System Simulator

Science.gov (United States)

Vranos, A.; Marteney, P. J.

1980-01-01

The thermal stability of aircraft gas turbines fuels was investigated. The objectives were: (1) to design and build an aircraft fuel system simulator; (2) to establish criteria for quantitative assessment of fuel thermal degradation; and (3) to measure the thermal degradation of Jet A and an alternative fuel. Accordingly, an aircraft fuel system simulator was built and the coking tendencies of Jet A and a model alternative fuel (No. 2 heating oil) were measured over a range of temperatures, pressures, flows, and fuel inlet conditions.

Effect of titania addition on the thermal conductivity of UO2 fuel [Paper IIIB-C

International Nuclear Information System (INIS)

Sengupta, A.K.; Kumar, A.; Arora, K.B.S.; Pandey, V.D.; Nair, M.R.; Kamath, H.S.

1986-01-01

Pellet clad interaction in nuclear reactor fuel elements can be reduced by the use of higher grain size UO 2 fuel. This is achieved by the addition of dopant like titania, niobia etc. However, these dopants are considered as impurities which may affect the thermophysical and thermomechanical properties of the fuel. Thermal Conductivity which is one of the important properties controlling the inpile performance of the fuel has been measured for pure UO 2 and UO 2 containing 0.05wt per cent and 0.1wt per cent TiO 2 in the temperature range 900K to 1900K in vacuum. Thermal conductivity was obtained from thermal diffusivity data measured by laser flash method. The paper highlights the experimental results and discusses the effect of TiO 2 on the thermal conductivity of UO 2 fuel. (author)

Effect of titania addition on the thermal conductivity of UO2 fuel (Paper IIIB-C)

Energy Technology Data Exchange (ETDEWEB)

Sengupta, A K; Kumar, A; Arora, K B.S.; Pandey, V D; Nair, M R; Kamath, H S

1986-01-01

Pellet clad interaction in nuclear reactor fuel elements can be reduced by the use of higher grain size UO2 fuel. This is achieved by the addition of dopant like titania, niobia etc. However, these dopants are considered as impurities which may affect the thermophysical and thermomechanical properties of the fuel. Thermal Conductivity which is one of the important properties controlling the inpile performance of the fuel has been measured for pure UO2 and UO2 containing 0.05wt per cent and 0.1wt per cent TiO2 in the temperature range 900K to 1900K in vacuum. Thermal conductivity was obtained from thermal diffusivity data measured by laser flash method. The paper highlights the experimental results and discusses the effect of TiO2 on the thermal conductivity of UO2 fuel. 5 figures.

Demonstration of Passive Fuel Cell Thermal Management Technology

Science.gov (United States)

Burke, Kenneth A.; Jakupca, Ian; Colozza, Anthony; Wynne, Robert; Miller, Michael; Meyer, Al; Smith, William

2012-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA Exploration program. The passive thermal management system relies on heat conduction within highly thermally conductive cooling plates to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack. Using the passive approach eliminates the need for a coolant pump and other cooling loop components within the fuel cell system which reduces mass and improves overall system reliability. Previous development demonstrated the performance of suitable highly thermally conductive cooling plates and integrated heat exchanger technology to collect the heat from the cooling plates (Ref. 1). The next step in the development of this passive thermal approach was the demonstration of the control of the heat removal process and the demonstration of the passive thermal control technology in actual fuel cell stacks. Tests were run with a simulated fuel cell stack passive thermal management system outfitted with passive cooling plates, an integrated heat exchanger and two types of cooling flow control valves. The tests were run to demonstrate the controllability of the passive thermal control approach. Finally, successful demonstrations of passive thermal control technology were conducted with fuel cell stacks from two fuel cell stack vendors.

Fuel thermal conductivity (FTHCON). Status report

International Nuclear Information System (INIS)

Hagrman, D.L.

1979-02-01

An improvement of the fuel thermal conductivity subcode is described which is part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The original version was published in the Materials Properties (MATPRO) Handbook, Section A-2 (Fuel Thermal Conductivity). The improved version incorporates data which were not included in the previous work and omits some previously used data which are believed to come from cracked specimens. The models for the effect of porosity on thermal conductivity and for the electronic contribution to thermal coductivity have been completely revised in order to place these models on a more mechanistic basis. As a result of modeling improvements the standard error of the model with respect to its data base has been significantly reduced

The efficiency and fidelity of the in-core nuclear fuel management code FORMOSA-P

International Nuclear Information System (INIS)

Kropaczek, D.J.; Turinsky, P.J.

1994-01-01

The second-order generalized perturbation theory (GPT), nodal neutronic model utilized within the nuclear fuel management optimization code FORMOSA-P is presented within the context of prediction fidelity and computational efficiency versus forward solution. Key features of thr GPT neutronics model as implemented within the Simulated Annealing optimization adaptive control algorithm are discussed. Supporting results are then presented demonstrating the superior consistency of adaptive control for both global and local optimization searches. (authors). 15 refs., 1 fig., 4 tabs

An Innovative High Thermal Conductivity Fuel Design

Energy Technology Data Exchange (ETDEWEB)

Jamil A. Khan

2009-11-21

Thermal conductivity of the fuel in today's Light Water Reactors, Uranium dioxide, can be improved by incorporating a uniformly distributed heat conducting network of a higher conductivity material, Silicon Carbide. The higher thermal conductivity of SiC along with its other prominent reactor-grade properties makes it a potential material to address some of the related issues when used in UO2 [97% TD]. This ongoing research, in collaboration with the University of Florida, aims to investigate the feasibility and develop a formal methodology of producing the resultant composite oxide fuel. Calculations of effective thermal conductivity of the new fuel as a function of %SiC for certain percentages and as a function of temperature are presented as a preliminary approach. The effective thermal conductivities are obtained at different temperatures from 600K to 1600K. The corresponding polynomial equations for the temperature-dependent thermal conductivities are given based on the simulation results. Heat transfer mechanism in this fuel is explained using a finite volume approach and validated against existing empirical models. FLUENT 6.1.22 was used for thermal conductivity calculations and to estimate reduction in centerline temperatures achievable within such a fuel rod. Later, computer codes COMBINE-PC and VENTURE-PC were deployed to estimate the fuel enrichment required, to maintain the same burnup levels, corresponding to a volume percent addition of SiC.

An Innovative High Thermal Conductivity Fuel Design

International Nuclear Information System (INIS)

Khan, Jamil A.

2009-01-01

Thermal conductivity of the fuel in today's Light Water Reactors, Uranium dioxide, can be improved by incorporating a uniformly distributed heat conducting network of a higher conductivity material, Silicon Carbide. The higher thermal conductivity of SiC along with its other prominent reactor-grade properties makes it a potential material to address some of the related issues when used in UO2 (97% TD). This ongoing research, in collaboration with the University of Florida, aims to investigate the feasibility and develop a formal methodology of producing the resultant composite oxide fuel. Calculations of effective thermal conductivity of the new fuel as a function of %SiC for certain percentages and as a function of temperature are presented as a preliminary approach. The effective thermal conductivities are obtained at different temperatures from 600K to 1600K. The corresponding polynomial equations for the temperature-dependent thermal conductivities are given based on the simulation results. Heat transfer mechanism in this fuel is explained using a finite volume approach and validated against existing empirical models. FLUENT 6.1.22 was used for thermal conductivity calculations and to estimate reduction in centerline temperatures achievable within such a fuel rod. Later, computer codes COMBINE-PC and VENTURE-PC were deployed to estimate the fuel enrichment required, to maintain the same burnup levels, corresponding to a volume percent addition of SiC.

Fuel Thermal Expansion (FTHEXP)

International Nuclear Information System (INIS)

Reymann, G.A.

1978-07-01

A model is presented which deals with dimensional changes in LWR fuel pellets caused by changes in temperature. It is capable of dealing with any combination of UO 2 and PuO 2 in solid, liquid or mixed phase states, and includes expansion due to the solid-liquid phase change. The function FTHEXP models fuel thermal expansion as a function of temperature, fraction of PuO 2 , and the fraction of fuel which is molten

Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

2002-01-01

Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

A recommendation for the thermal conductivity of oxide fuels

International Nuclear Information System (INIS)

Kang, K. H.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Na, S. H.; Lee, Y. W.; Moon, H. S.; Kim, H. S.

2004-01-01

The thermal conductivity of nuclear fuel is one of the most important properties because it affects the fuel operating temperature. Therefore, it influences almost all the important processes occurred in nuclear fuel during irradiation, such as gas release, swelling and grain growth. The model of the thermal conductivity of nuclear fuel should be used in the codes to evaluate the performance of it analytically and be required in the nuclear fuel research and development. The thermal conductivity, k, of UO 2 depends on the deviation from stoichiometry, x, the burnup, b, and the fractional porosity, p, as well as the temperature, T: k = k(x, b, p, T), (1) Changes in thermal conductivity occur during irradiation because of fission-gas bubble formation, pores, cracks, fission product build-up and possible changes in the oxygen to uranium ratio (O/U). The dependence on temperature and porosity has been well studied and incorporated in computer codes used for the in-pile fuel behavior analysis. There are several studies on the effect of impurity on the thermal conductivity of UO 2 . In this paper, the variables affected on the thermal conductivity were studied. The available data of the thermal conductivity of UO 2 , UO 2+x , (U, Pu)O 2 , (U, Pu)O 2 and simulated fuel for irradiation fuel were reviewed and analyzed. The best models were recommended

System evaluation of improved thermal stability jet fuels

Energy Technology Data Exchange (ETDEWEB)

Binns, K.E.; Dieterle, G.L.; Williams, T. [Univ. of Dayton Research Institute, OH (United States)

1995-05-01

A single-pass, single-tube heat exchanger device called the Phoenix rig and a single-pass, dual-heat exchanger system called the Extended Duration Thermal Stability Test system are specific devices/systems developed for evaluating jet fuel thermal stability. They have been used extensively in the evaluation of various jet fuels and thermal stability additives. The test results have indicated that additives can substantially improve the thermal stability of conventional jet fuels. Relationships of oxygen consumption, residence time, bulk, and wetted wall temperatures on coking deposits that form in the heated tubes have also been investigated.

Thermal model of spent fuel transport cask

International Nuclear Information System (INIS)

Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

1996-01-01

The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

Quantum Fidelity and Thermal Phase Transitions in a Two-Dimensional Spin System

International Nuclear Information System (INIS)

Wang Bo; Kou Su-Peng; Huang Hai-Lin; Sun Zhao-Yu

2012-01-01

We investigate the ability of quantum fidelity in detecting the classical phase transitions (CPTs) in a two-dimensional Heisenberg—Ising mixed spin model, which has a very rich phase diagram and is exactly soluble. For a two-site subsystem of the model, the reduced fidelity (including the operator fidelity and the fidelity susceptibility) at finite temperatures is calculated, and it is found that an extreme value presents at the critical temperature, thus shows a signal for the CPTs. In some parameter region, the signal becomes blurred. We propose to use the 'normalized fidelity susceptibility' to solve this problem

Thermal Cycling of Uranium Dioxide - Tungsten Cermet Fuel Specimens

Energy Technology Data Exchange (ETDEWEB)

Gripshover, P.J.; Peterson, J.H.

1969-12-08

In phase I tungsten clad cermet fuel specimens were thermal cycled, to study the effects of fuel loading, fuel particle size, stablized fuel, duplex coatings, and fabrication techniques on dimensional stability during thermal cycling. In phase II the best combination of the factors studies in phase I were combined in one specimen for evaluation.

Thermally-induced bowing of CANDU fuel elements

International Nuclear Information System (INIS)

Suk, H.C.; Sim, K.S.; Park, J.H.; Park, G.S.

1995-01-01

Considering only the thermally-induced bending moments which are generated both within the sheath and between the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element, a generalized and explicit analytical formula for the thermally-induced bending is developed in this paper, based on the cases of 1) the bending of an empty tube treated by neglecting of the fuel/sheath mechanical interaction and 2) the fuel/sheath interaction due to the pellet and sheath temperature variations. In each of the cases, the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. Investigating the relative importance of the various parameters affecting fuel element bowing, the element bowing is found to be greatly affected with the variations of element length, sheath diameter, pellet/sheath mechanical interaction and neutron flux depression factors, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient, and sheath and pellet thermal conductivities. Also, the element bowing of the standard 37-element bundle and CANFLEX 43-element bundle for the use in CANDU-6 reactors was analyzed with the formula, which could help to demonstrate the integrity of the fuel. All the required input data for the analyses were generated in terms of the reactor operation conditions on the reactor physics, thermal hydraulics and fuel performance by using various CANDU computer codes. The analysis results indicate that the CANFLEX 43-element

Thermal neutron spectrum distribution in TRIGA fuels

International Nuclear Information System (INIS)

Gui Ah Auu; Harasawa, Susumu; An, Shigehiro

1989-01-01

The dependence of thermal neutron spectrum in TRIGA fuel cell on fuel temperature and TRIGA fuel types were studied using LIBP and THERMOS codes. Some characteristics of the TRIGA fuel including its prompt negative temperature coefficient of reactivity were explained using the results of the study. (author)

Jet Fuel Thermal Stability Investigations Using Ellipsometry

Science.gov (United States)

Nash, Leigh; Vasu, Subith S.; Klettlinger, Jennifer Lindsey

2017-01-01

Jet fuels are typically used for endothermic cooling in practical engines where their thermal stability is very important. In this work the thermal stability of Sasol IPK (a synthetic jet fuel) with varying levels of naphthalene has been studied on stainless steel substrates using spectroscopic ellipsometry in the temperature range 385-400 K. Ellipsometry is an optical technique that measures the changes in a light beam’s polarization and intensity after it reflects off of a thin film to determine the film’s thickness and optical properties. All of the tubes used were rated as thermally unstable by the color standard portion of the Jet Fuel Thermal Oxidation Test, and this was confirmed by the deposit thicknesses observed using ellipsometry. A new amorphous model on a stainless steel substrate was used to model the data and obtain the results. It was observed that, as would be expected, increasing the temperature of the tube increased the overall deposit amount for a constant concentration of naphthalene. The repeatability of these measurements was assessed using multiple trials of the same fuel at 385 K. Lastly, the effect of increasing the naphthalene concentration in the fuel at a constant temperature was found to increase the deposit thickness.In conclusion, ellipsometry was used to investigate the thermal stability of jet fuels on stainless steel substrate. The effects of increasing temperature and addition of naphthalene on stainless steel tubes with Sasol IPK fuel were investigated. It was found, as expected, that increasing temperature lead to an increase in deposit thickness. It wasAmerican Institute of Aeronautics and Astronautics6also found that increasing amounts of naphthalene increased the maximum deposit thickness. The repeatability of these measurements was investigated using multiple tests at the same conditions. The present work provides as a better quantitative tool compared to the widely used JFTOT technique. Future work will expand on the

Thermal breeder fuel enrichment zoning

International Nuclear Information System (INIS)

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

1992-01-01

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

Thermal Analysis Evaluation of Spent Fuel Storage Rack for Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Sangjin; Oh, Jinho; Kwak, Jinsung; Lee, Jongmin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Spent fuel storage rack is to store spent fuel assemblies. The spent fuel storage rack is submerged in the designated pool for cooling. Due to the condition change of the pool water, the effect of thermal load on spent fuel storage rack must be analyzed and evaluated. In this paper, thermal stress analysis is performed and evaluated on a spent fuel storage rack. For thermal stress evaluation of the spent fuel storage rack, load combinations and allowable criteria in ASME Sec. III NB-3220 are applied. In cases of A-1 and B-1, the same temperature applied on the whole model, thermal stress doesn't occur because there is no constraint about the thermal expansion. The support frame is located on the pool bottom in free standing type and the racks are located in the support frame with enough space. Thermal expansion was considered and reflected in the design of spent fuel storage rack in advance. Thermal stress analysis is performed and evaluated on a spent fuel storage rack with consideration of pool water temperature variation. The thermal analysis including a linear heat transfer and the thermal stress analysis is performed for the racks and support frame and resulted stresses are within allowable criteria.

Fuel thermal conductivity (FTHCON). Status report. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Hagrman, D. L.

1979-02-01

An improvement of the fuel thermal conductivity subcode is described which is part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The original version was published in the Materials Properties (MATPRO) Handbook, Section A-2 (Fuel Thermal Conductivity). The improved version incorporates data which were not included in the previous work and omits some previously used data which are believed to come from cracked specimens. The models for the effect of porosity on thermal conductivity and for the electronic contribution to thermal coductivity have been completely revised in order to place these models on a more mechanistic basis. As a result of modeling improvements the standard error of the model with respect to its data base has been significantly reduced.

A study on the thermal expansion characteristics of simulated spent fuel and simulated DUPIC fuel

International Nuclear Information System (INIS)

Kang, Kweon Ho; Ryu, H. J.; Kim, H. S.; Song, K. C.; Yang, M. S.

2001-10-01

Thermal expansions of simulated spent PWR fuel and simulated DUPIC fuel were studied using a dilatometer in the temperature range from 298 to 1900 K. The densities of simulated spent PWR fuel and simulated DUPIC fuel used in the measurement were 10.28 g/cm3 (95.35 % of TD) and 10.26 g/cm3 (95.14 % of TD), respectively. Their linear thermal expansions of simulated fuels are higher than that of UO2, and the difference between these fuels and UO2 increases progressively as temperature increases. However, the difference between simulated spent PWR fuel and simulated DUPIC fuel can hardly be observed. For the temperature range from 298 to 1900 K, the values of the average linear thermal expansion coefficients for simulated spent PWR fuel and simulated DUPIC fuel are 1.391 10-5 and 1.393 10-5 K-1, respectively. As temperature increases to 1900 K, the relative densities of simulated spent PWR fuel and simulated DUPIC fuel decrease to 93.81 and 93.76 % of initial densities at 298 K, respectively

Thermal expansion of UO2 and simulated DUPIC fuel

International Nuclear Information System (INIS)

Ho Kang, Kweon; Jin Ryu, Ho; Chan Song, Kee; Seung Yang, Myung

2002-01-01

The lattice parameters of simulated DUPIC fuel and UO 2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO 2 , and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO 2 . For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO 2 and simulated DUPIC fuel are 10.471x10 -6 and 10.751x10 -6 K -1 , respectively

Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

International Nuclear Information System (INIS)

Umbehaun, Pedro Ernesto

2000-01-01

This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

Thermal insulation of fuel elements

International Nuclear Information System (INIS)

Dubrovcak, P.; Pec, V.; Pitonak, J.

1978-01-01

The claim of the invention concerns thermal insulation of fuel elements heated for measurement of uranium fuel physical properties. For this, layers of aluminium film and of glass fibre are wound onto the inner tube of the element cladding. The space between the inner and the outer tubes is evacuated and the tubes are spaced using spacer wires. (M.S.)

Thermal Radiation Effects on Thermal Explosion in Polydisperse Fuel Spray-Probabilistic Model

Directory of Open Access Journals (Sweden)

Ophir Navea

2011-06-01

Full Text Available We investigate the effect of thermal radiation on the dynamics of a thermal explosion of polydisperse fuel spray with a complete description of the chemistry via a single-step two-reactant model of general order. The polydisperse spray is modeled using a Probability Density Function (PDF. The thermal radiation energy exchange between the evaporation surface of the fuel droplets and the burning gas is described using the Marshak boundary conditions. An explicit expression of the critical condition for thermal explosion limit is derived analytically and represents a generalization of the critical parameter of the classical Semenov theory. Because we investigated the model in the range where the temperature is very high, the effect of the thermal radiation is significant.

Experimental determination of fuel-cladding thermal contact resistance

International Nuclear Information System (INIS)

Maglic, K.; Zivotic, Z.

1968-01-01

Thermal resistance of the UO 2 fuel - Zr-2 cladding was measure by the same experimental apparatus which was used for measuring the thermal conductivity of ceramic fuel. Thermal resistance was measure for a series of heat flux values and the dependence of thermal resistance on the flux is given within in the range from 0.66 W/cm 2 to 13.3 W/cm 2 . The temperature drop on the contact surface was between 39 deg C and 181.7 deg C, proportional to the increase of the heat flux [sr

Thermal-hydraulic analysis of spent fuel storage systems

International Nuclear Information System (INIS)

Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

1987-01-01

This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs

Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Choi, Hang Bok

2005-03-01

Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

International Nuclear Information System (INIS)

Mishima, Kaichiro; Shibata, Toshikazu.

1982-01-01

This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

Thermal phenomenae in nuclear fuel rods

International Nuclear Information System (INIS)

Baigorria, Carlos.

1983-12-01

Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor) [es

Advanced Fuel Cell System Thermal Management for NASA Exploration Missions

Science.gov (United States)

Burke, Kenneth A.

2009-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA exploration program. An analysis of a state-of-the-art fuel cell cooling systems was done to benchmark the portion of a fuel cell system s mass that is dedicated to thermal management. Additional analysis was done to determine the key performance targets of the advanced passive thermal management technology that would substantially reduce fuel cell system mass.

Improving the Thermal Conductivity of UO2 Fuel with the Addition of Graphene

International Nuclear Information System (INIS)

Cho, Byoung Jin; Kim, Young Jin; Sohn, Dong Seong

2012-01-01

Improvement of fuel performances by increasing the fuel thermal conductivity using the BeO or W were reported elsewhere. In this paper, some major fuel performances of improved thermal conductivity oxide (ICO) nuclear fuel with the addition of 10 v/o graphene have been compared to those of standard UO 2 fuel. The fuel thermal conductivity affects many performance parameters and thus is an important parameter to determine the fuel performance. Furthermore, it also affects the performance of the fuel during reactor accidents. The improved thermal conductivity of the fuel would reduce the fuel temperature at the same power condition and would improve the fission gas release, rod internal pressure and fuel stored energy. Graphene is well known for its excellent electrical conductivity, strength and thermal conductivity. The addition of graphene to the UO 2 fuel could increase the thermal conductivity of the ICO fuel. Although the graphene material is extensively studied recently, the characteristics of the graphene material, especially the thermal properties, are not well-known yet. In this study, we used the Light Water Reactor fuel performance analysis code FRAPCON-3.2 to analyze the performance of standard UO 2 and ICO fuel

Development of Passive Fuel Cell Thermal Management Heat Exchanger

Science.gov (United States)

Burke, Kenneth A.; Jakupca, Ian J.; Colozza, Anthony J.

2010-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA Exploration program. The passive thermal management system relies on heat conduction within highly thermally conductive cooling plates to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack. Using the passive approach eliminates the need for a coolant pump and other cooling loop components within the fuel cell system which reduces mass and improves overall system reliability. Previous development demonstrated the performance of suitable highly thermally conductive cooling plates that could conduct the heat, provide a sufficiently uniform temperature heat sink for each cell of the fuel cell stack, and be substantially lighter than the conventional thermal management approach. Tests were run with different materials to evaluate the design approach to a heat exchanger that could interface with the edges of the passive cooling plates. Measurements were made during fuel cell operation to determine the temperature of individual cooling plates and also to determine the temperature uniformity from one cooling plate to another.

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

International Nuclear Information System (INIS)

Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

2011-01-01

The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

Transmutation Fuel Performance Code Thermal Model Verification

Energy Technology Data Exchange (ETDEWEB)

Gregory K. Miller; Pavel G. Medvedev

2007-09-01

FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

Maximum thermal loading test of BWR fuel assembly

International Nuclear Information System (INIS)

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

1987-01-01

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

Toward Improved Fidelity of Thermal Explosion Simulations

Energy Technology Data Exchange (ETDEWEB)

Nichols, A L; Becker, R; Howard, W M; Wemhoff, A

2009-07-17

We will present results of an effort to improve the thermal/chemical/mechanical modeling of HMX based explosive like LX04 and LX10 for thermal cook-off. The original HMX model and analysis scheme were developed by Yoh et.al. for use in the ALE3D modeling framework. The current results were built to remedy the deficiencies of that original model. We concentrated our efforts in four areas. The first area was addition of porosity to the chemical material model framework in ALE3D that is used to model the HMX explosive formulation. This is needed to handle the roughly 2% porosity in solid explosives. The second area was the improvement of the HMX reaction network, which included the inclusion of a reactive phase change model base on work by Henson et.al. The third area required adding early decomposition gas species to the CHEETAH material database to develop more accurate equations of state for gaseous intermediates and products. Finally, it was necessary to improve the implicit mechanics module in ALE3D to more naturally handle the long time scales associated with thermal cook-off. The application of the resulting framework to the analysis of the Scaled Thermal Explosion (STEX) experiments will be discussed.

Azobenzene-functionalized carbon nanotubes as high-energy density solar thermal fuels.

Science.gov (United States)

Kolpak, Alexie M; Grossman, Jeffrey C

2011-08-10

Solar thermal fuels, which reversibly store solar energy in molecular bonds, are a tantalizing prospect for clean, renewable, and transportable energy conversion/storage. However, large-scale adoption requires enhanced energy storage capacity and thermal stability. Here we present a novel solar thermal fuel, composed of azobenzene-functionalized carbon nanotubes, with the volumetric energy density of Li-ion batteries. Our work also demonstrates that the inclusion of nanoscale templates is an effective strategy for design of highly cyclable, thermally stable, and energy-dense solar thermal fuels.

Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2016-05-15

U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development

International Nuclear Information System (INIS)

Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

2007-01-01

The uranium dioxide is the most used substance as nuclear reactor fuel for presenting many advantages such as: high stability even when it is in contact with water in high temperatures, high fusion point, and high capacity to retain fission products. The conventional fuel is made with ceramic sintered pellets of uranium dioxide stacked inside fuel rods, and presents disadvantages because its low thermal conductivity causes large and dangerous temperature gradients. Besides, the thermal conductivity decreases further as the fuel burns, what limits a pellet operational lifetime. This research developed a new kind of fuel pellets fabricated with uranium dioxide kernels and beryllium oxide filling the empty spaces between them. This fuel has a great advantage because of its higher thermal conductivity in relation to the conventional fuel. Pellets of this kind were produced, and had their thermophysical properties measured by the flash laser method, to compare with the thermal conductivity of the conventional uranium dioxide nuclear fuel. (author) (author)

Thermal expansion study of simulated DUPIC fuel using neutron diffraction

International Nuclear Information System (INIS)

Kang, Kweon Ho; Ryu, H. J.; Bae, J. H.; Kim, H. S.; Song, K. C.; Yang, M. S.; Choi, Y. N.; Han, Y. S.; Oh, H. S.

2001-07-01

The lattice parameters of simulated DUPIC fuel and UO2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO2 and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO2. For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO2 and simulated DUPIC fuel are 10.471 ''10-6 and 10.751 ''10-6 K-1, respectively

Lumped-parameter fuel rod model for rapid thermal transients

International Nuclear Information System (INIS)

Perkins, K.R.; Ramshaw, J.D.

1975-07-01

The thermal behavior of fuel rods during simulated accident conditions is extremely sensitive to the heat transfer coefficient which is, in turn, very sensitive to the cladding surface temperature and the fluid conditions. The development of a semianalytical, lumped-parameter fuel rod model which is intended to provide accurate calculations, in a minimum amount of computer time, of the thermal response of fuel rods during a simulated loss-of-coolant accident is described. The results show good agreement with calculations from a comprehensive fuel-rod code (FRAP-T) currently in use at Aerojet Nuclear Company

In-pile measurement of the thermal conductivity of irradiated metallic fuel

International Nuclear Information System (INIS)

Bauer, T.H.; Holland, J.W.

1995-01-01

Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

Thermal bonding of light water reactor fuel using nonalkaline liquid-metal alloy

International Nuclear Information System (INIS)

Wright, R.F.; Tulenko, J.S.; Schoessow, G.J.; Connell, R.G. Jr.; Dubecky, M.A.; Adams, T.

1996-01-01

Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. A technique is explored that extends fuel performance by thermally bonding LWR fuel with a nonalkaline liquid-metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Because of the low thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high-conductivity liquid metal thermally bonds the fuel to the cladding and eliminates the large temperature change across the gap while preserving the expansion and pellet-loading capabilities. The application of liquid-bonding techniques to LWR fuel is explored to increase LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) is developed to analyze the in-reactor performance of the liquid-metal-bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of liquid-bonded LWR fuel. The results show that liquid-bonded boiling water reactor peak fuel temperatures are 400 F lower at beginning of life and 200 F lower at end of life compared with conventional fuel

Coupled Tort-TD/CTF Capability for high-fidelity LWR core calculations - 321

International Nuclear Information System (INIS)

Christienne, M.; Avramova, M.; Perin, Y.; Seubert, A.

2010-01-01

This paper describes the developed coupling scheme between TORT-TD and CTF. TORT-TD is a time-dependent 3D discrete ordinates neutron transport code. TORT-TD is utilized for high-fidelity reactor core neutronics calculations while CTF is providing the thermal-hydraulics feedback information. CTF is an improved version of the advanced thermal-hydraulic sub-channel code COBRA-TF, which is widely used for best-estimate evaluations of LWR safety margins. CTF is a transient code based on a separated flow representation of the two-phase flow. The coupled code TORT-TD/CTF allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. Steady-state and transient test cases, based on the OECD/NRC PWR MOX/UO 2 Core Transient Benchmark, have been calculated. The steady state cases are based on a quarter core model while the transient test case models a control rod ejection transient in a small PWR mini-core fuel assembly arrangement. The obtained results with TORT-TD/CTF are verified by a code-to-code comparison with the previously developed NEM/CTF and TORT-TD/ATHLET coupled code systems. The performed comparative analysis indicates the applicability and high-fidelity potential of the TORT-TD/CTF coupling. (authors)

Discrete Modeling of Early-Life Thermal Fracture in Ceramic Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Spencer, Benjamin W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Huang, Hai [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dolbow, John E. [Duke Univ., Durham, NC (United States); Hales, Jason D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-01

Fracturing of ceramic fuel pellets heavily influences performance of light water reactor (LWR) fuel. Early in the life of fuel, starting with the initial power ramp, large thermal gradients cause high tensile hoop and axial stresses in the outer region of the fuel pellets, resulting in the formation of radial and axial cracks. Circumferential cracks form due to thermal gradients that occur when the power is ramped down. These thermal cracks cause the fuel to expand radially, closing the pellet/cladding gap and enhancing the thermal conductance across that gap, while decreasing the effective conductivity of the fuel in directions normal to the cracking. At lower length scales, formation of microcracks is an important contributor to the decrease in bulk thermal conductivity that occurs over the life of the fuel as the burnup increases. Because of the important effects that fracture has on fuel performance, a realistic, physically based fracture modeling capability is essential to predict fuel behavior in a wide variety of normal and abnormal conditions. Modeling fracture within the context of the finite element method, which is based on continuous interpolations of solution variables, has always been challenging because fracture is an inherently discontinuous phenomenon. Work is underway at Idaho National Laboratory to apply two modeling techniques model fracture as a discrete displacement discontinuity to nuclear fuel: The extended finite element method (XFEM), and discrete element method (DEM). XFEM is based on the standard finite element method, but with enhancements to represent discontinuous behavior. DEM represents a solid as a network of particles connected by bonds, which can arbitrarily fail if a fracture criterion is reached. This paper presents initial results applying the aforementioned techniques to model fuel fracturing. This work has initially focused on early life behavior of ceramic LWR fuel. A coupled thermal-mechanical XFEM method that includes

Nuclear fuel element

International Nuclear Information System (INIS)

Yamamoto, Seigoro.

1994-01-01

Ultrafine particles of a thermal neutron absorber showing ultraplasticity is dispersed in oxide ceramic fuels by more than 1% to 10% or lower. The ultrafine particles of the thermal neutron absorber showing ultrafine plasticity is selected from any one of ZrGd, HfEu, HfY, HfGd, ZrEu, and ZrY. The thermal neutron absorber is converted into ultrafine particles and solid-solubilized in a nuclear fuel pellet, so that the dispersion thereof into nuclear fuels is made uniform and an absorbing performance of the thermal neutrons is also made uniform. Moreover, the characteristics thereof, for example, physical properties such as expansion coefficient and thermal conductivity of the nuclear fuels are also improved. The neutron absorber, such as ZrGd or the like, can provide plasticity of nuclear fuels, if it is mixed into the nuclear fuels for showing the plasticity. The nuclear fuel pellets are deformed like an hour glass as burning, but, since the end portion thereof is deformed plastically within a range of a repulsive force of the cladding tube, there is no worry of damaging a portion of the cladding tube. (N.H.)

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

International Nuclear Information System (INIS)

Liu, X.J.; Cheng, X.

2009-01-01

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

Irradiation effects on thermal properties of LWR hydride fuel

Energy Technology Data Exchange (ETDEWEB)

Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

2017-04-01

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

High Fidelity Simulation of Atomization in Diesel Engine Sprays

Science.gov (United States)

2015-09-01

state Figure 5. Q criterion isosurface colored by streamwise velocity in the diesel spray injector as viewed from the nozzle exit. Figure 6. U contour...fidelity simulation approach was adopted to study the atom- ization physics of a diesel injector with detailed nozzle internal geometry. The nozzle flow...26; Stanford, CA 14. ABSTRACT A high fidelity numerical simulation of jet breakup and spray formation from a complex diesel fuel injector has been

Thermophysical properties of the products of low-grade fuels thermal recycling

Directory of Open Access Journals (Sweden)

Tabakaev Roman B.

2015-01-01

Full Text Available The relevance of the work is caused by reorientation of the modern power engineering to use of local low grade fuel resources. Some types of low grade fuels (peat, brown coal, sapropel, wood chips are considered in this work. Thermotechnical characteristics of the investigated fuels and products of their thermal recycling are determined. Thermal recycling process is accompanied by release of fuel dissociation heat (0.33-3.69 MJ/kg. The results of thermal low grade fuel recycling are solid carbonaceous product (semi-coke with a calorific value higher in 1.5-7 times than the value of natural fuels; pyrolysis resin with calorific value 29.4-36.8 MJ/kg; combustible gas with calorific value 15.16-19.06 MJ/m3.

GAPCON-THERMAL-2: a computer program for calculating the thermal behavior of an oxide fuel rod

International Nuclear Information System (INIS)

Beyer, C.E.; Hann, C.R.; Lanning, D.D.; Panisko, F.E.; Parchen, L.J.

1975-11-01

A description is presented of the computer code GAPCON THERMAL-2, a light water reactor (LWR) fuel thermal performance prediction code. GAPCON-THERMAL-2, is intended to be used as a calculational tool for reactor fuel steady-state thermal performance and to provide input for accident analyses. Some models used in the code provide best estimate as well as conservative predictions. Each of the individual models in the code is based on the best available data

Evaluation of thermal physical properties for fast reactor fuels. Melting point and thermal conductivities

International Nuclear Information System (INIS)

Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki; Ogasawara, Masahiro; Tamura, Tetsuya; Sugata, Hirotada; Sunaoshi, Takeo; Shibata, Kazuya

2006-10-01

Japan Atomic Energy Agency has developed a fast breeder reactor (FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio (O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Neumann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content. (author)

Characterization of a novel, highly integrated tubular solid oxide fuel cell system using high-fidelity simulation tools

Science.gov (United States)

Kattke, K. J.; Braun, R. J.

2011-08-01

A novel, highly integrated tubular SOFC system intended for small-scale power is characterized through a series of sensitivity analyses and parametric studies using a previously developed high-fidelity simulation tool. The high-fidelity tubular SOFC system modeling tool is utilized to simulate system-wide performance and capture the thermofluidic coupling between system components. Stack performance prediction is based on 66 anode-supported tubular cells individually evaluated with a 1-D electrochemical cell model coupled to a 3-D computational fluid dynamics model of the cell surroundings. Radiation is the dominate stack cooling mechanism accounting for 66-92% of total heat loss at the outer surface of all cells at baseline conditions. An average temperature difference of nearly 125 °C provides a large driving force for radiation heat transfer from the stack to the cylindrical enclosure surrounding the tube bundle. Consequently, cell power and voltage disparities within the stack are largely a function of the radiation view factor from an individual tube to the surrounding stack can wall. The cells which are connected in electrical series, vary in power from 7.6 to 10.8 W (with a standard deviation, σ = 1.2 W) and cell voltage varies from 0.52 to 0.73 V (with σ = 81 mV) at the simulation baseline conditions. It is observed that high cell voltage and power outputs directly correspond to tubular cells with the smallest radiation view factor to the enclosure wall, and vice versa for tubes exhibiting low performance. Results also reveal effective control variables and operating strategies along with an improved understanding of the effect that design modifications have on system performance. By decreasing the air flowrate into the system by 10%, the stack can wall temperature increases by about 6% which increases the minimum cell voltage to 0.62 V and reduces deviations in cell power and voltage by 31%. A low baseline fuel utilization is increased by decreasing the

Behavior of mixed-oxide fuel subjected to multiple thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Hofman, G.L.; Neimark, L.A.; Poeppel, R.B.

1983-11-01

The microstructural behavior of irradiated mixed-oxide fuel subjected to multiple, mild thermal transients was investigated using direct electrical heating. The results demonstrate that significant intergranular porosity, accompanied by large-scale (>90%) release of the retained fission gas, developed as a result of the cyclic heating. Microstructural examination of the fuel indicated that thermal-shock-induced cracking of the fuel contributed significantly to the increased swelling and gas release

Thermal ramp rate effects on mixed-oxide fuel swelling/gas release

International Nuclear Information System (INIS)

Hinman, C.A.; Randklev, E.H.

1979-01-01

Macroscopic swelling behavior of PNL-10 was compared to that of PNL-2 fuel and it was found that the swelling-threshold behavior is similar for similar thermal conditions. Transient fission gas release for the PNL-10 fuel is very similar to that observed for the PNL-2 fuel for similar thermal conditions

Thermal conductivity of heterogeneous LWR MOX fuels

Science.gov (United States)

Staicu, D.; Barker, M.

2013-11-01

It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

Compatibility analysis of DUPIC fuel(4) - thermal hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jee Won; Chae, Kyung Myung; Choi, Hang Bok

2000-07-01

Thermal-hydraulic compatibility of the DUPIC fuel bundle in the CANDU reactor has been studied. The critical channel power, the critical power ratio, the channel exit quality and the channel flow are calculated for the DUPIC and the standard fuels by using the NUCIRC code. The physical models and associated parametric values for the NUCIRC analysis of the fuels are also presented. Based upon the slave channel analysis, the critical channel power and the critical power ratios have been found to be very similar for the two fuel types. The same dryout model is used in this study for the standard and the DUPIC fuel bundles. To assess the dryout characteristics of the DUPIC fuel bundle, the ASSERT-PV code has been used for the subchannel analysis. Based upon the results of the subchannel analysis, it is found that the dryout location and the power for the two fuel types are indeed very similar. This study shows that thermal performance of the DUPIC fuel is not significantly different from that of the standard fuel.

Behavior of mixed-oxide fuel subjected to multiple thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Neimark, L.A.; Poeppel, R.B.; Hofman, G.L.

1985-01-01

The microstructural behavior of irradiated mixed-oxide fuel subjected to multiple, mild thermal transients was investigated using direct electrical heating. The results demonstrate that significant intergranular porosity, accompanied by large-scale (>90%) release of the retained fission gas, developed as a result of the cyclic heating. Microstructural examination of the fuel indicated that thermal-shock-induced cracking of the fuel contributed significantly to the increased swelling and gas release. 29 refs., 12 figs

Advanced thermally stable jet fuels: Technical progress report, October 1994--December 1994

Energy Technology Data Exchange (ETDEWEB)

Schobert, H.H.; Eser, S.; Song, C.; Hatcher, P.G.; Boehman, A.; Coleman, M.M.

1995-02-01

There are five tasks within this project on thermally stable coal-based jet fuels. Progress on each of the tasks is described. Task 1, Investigation of the quantitative degradation chemistry of fuels, has 5 subtasks which are described: Literature review on thermal stability of jet fuels; Pyrolytic and catalytic reactions of potential endothermic fuels: cis- and trans-decalin; Use of site specific {sup 13}C-labeling to examine the thermal stressing of 1-phenylhexane: A case study for the determination of reaction kinetics in complex fuel mixtures versus model compound studies; Estimation of critical temperatures of jet fuels; and Surface effects on deposit formation in a flow reactor system. Under Task 2, Investigation of incipient deposition, the subtask reported is Uncertainty analysis on growth and deposition of particles during heating of coal-derived aviation gas turbine fuels; under Task 3, Characterization of solid gums, sediments, and carbonaceous deposits, is subtask, Studies of surface chemistry of PX-21 activated carbon during thermal degradation of jet A-1 fuel and n-dodecane; under Task 4, Coal-based fuel stabilization studies, is subtask, Exploratory screening and development potential of jet fuel thermal stabilizers over 400 C; and under Task 5, Exploratory studies on the direct conversion of coal to high quality jet fuels, are 4 subtasks: Novel approaches to low-severity coal liquefaction and coal/resid co-processing using water and dispersed catalysts; Shape-selective naphthalene hydrogenation for production of thermally stable jet fuels; Design of a batch mode and a continuous mode three-phase reactor system for the liquefaction of coal and upgrading of coal liquids; and Exploratory studies on coal liquids upgrading using mesopores molecular sieve catalysts. 136 refs., 69 figs., 24 tabs.

TRANSPA: a code for transient thermal analysis of a single fuel pin

International Nuclear Information System (INIS)

Prenger, F.C.

1985-02-01

An analytical model (TRANSPA) for the transient thermal analysis of a single uranium carbide fuel pin was developed. This model uses thermal boundary conditions obtained from COBRA-WC output and calculates the transient thermal response of a single fuel pin to changes in internal power generation, coolant flowrate, or fuel pin physical configuration. The model uses the MITAS finite difference thermal analyzer. MITAS provides the means to input separate conductance models through the use of a user subroutine input capability. The model is a lumped-mass representation of the fuel pin using 26 nodes and 42 conductors. Run time for each transient analysis is approximately one minute of central processor time on the NOS operating system

Autoxidation of jet fuels: Implications for modeling and thermal stability

Energy Technology Data Exchange (ETDEWEB)

Heneghan, S.P. [Univ. of Dayton Research Institute, OH (United States); Chin, L.P. [Systems Research Laboratories, Inc., Dayton, OH (United States)

1995-05-01

The study and modeling of jet fuel thermal deposition is dependent on an understanding of and ability to model the oxidation chemistry. Global modeling of jet fuel oxidation is complicated by several facts. First, liquid jet fuels are hard to heat rapidly and fuels may begin to oxidize during the heat-up phase. Non-isothermal conditions can be accounted for but the evaluation of temperature versus time is difficult. Second, the jet fuels are a mixture of many compounds that may oxidize at different rates. Third, jet fuel oxidation may be autoaccelerating through the decomposition of the oxidation products. Attempts to model the deposition of jet fuels in two different flowing systems showed the inadequacy of a simple two-parameter global Arrhenius oxidation rate constant. Discarding previous assumptions about the form of the global rate constants results in a four parameter model (which accounts for autoacceleration). This paper discusses the source of the rate constant form and the meaning of each parameter. One of these parameters is associated with the pre-exponential of the autoxidation chain length. This value is expected to vary inversely to thermal stability. We calculate the parameters for two different fuels and discuss the implication to thermal and oxidative stability of the fuels. Finally, we discuss the effect of non-Arrhenius behavior on current modeling of deposition efforts.

The influence of porosity on the thermal conductivity of irradiated UO2 fuel

International Nuclear Information System (INIS)

Bakker, K.; Kwast, H.; Cordfunke, E.H.P.

1994-12-01

The influence of porosity on the thermal conductivity of irradiated UO 2 fuel has been determined with the Finite Element Method (FEM). Light-microscopy photographs were made of the fuel. The pore shape and the pore distribution are entered in the FEM program from these photographs. The two dimensional (2D) thermal conductivity in the plane of the photograph is obtained from the FEM calculations. The 2D thermal conductivity, that has no physical meaning itself, is the lower limit of the three dimensional (3D) thermal conductivity. For three well defined pore shapes the relation is determined between the 2D thermal conductivity and the 3D thermal conductivity. From these computations a simple relation is obtained that transfers the 2D thermal conductivity into the 3D thermal conductivity, independent of the pore shape. The influence of porosity on the 3D thermal conductivity of irradiated UO 2 fuel and UO 2 fuel doped with Nb 2 O 5 was computed with the FEM. (orig.)

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

Science.gov (United States)

Emrich, William

2009-01-01

An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

COMTA - a computer code for fuel mechanical and thermal analysis

International Nuclear Information System (INIS)

Basu, S.; Sawhney, S.S.; Anand, A.K.; Anantharaman, K.; Mehta, S.K.

1979-01-01

COMTA is a generalized computer code for integrity analysis of the free standing fuel cladding, with natural UO 2 or mixed oxide fuel pellets. Thermal and Mechanical analysis is done simultaneously for any power history of the fuel pin. For analysis, the fuel cladding is assumed to be axisymmetric and is subjected to axisymmetric load due to contact pressure, gas pressure, coolant pressure and thermal loads. Axial variation of load is neglected and creep and plasticity are assumed to occur at constant volume. The pellet is assumed to be made of concentric annuli. The fission gas release integral is dependent on the temperature and the power produced in each annulus. To calculate the temperature distribution in the fuel pin, the variation of bulk coolant temperature is given as an input to the code. Gap conductance is calculated at every time step, considering fuel densification, fuel relocation and gap closure, filler gas dilution by released fission gas, gap closure by expansion and irradiation swelling. Overall gap conductance is contributed by heat transfer due to the three modes; conduction convection and radiation as per modified Ross and Stoute model. Equilibrium equations, compatibility equations, stress strain relationships (including thermal strains and permanent strains due to creep and plasticity) are used to obtain triaxial stresses and strains. Thermal strain is assumed to be zero at hot zero power conditions. The boundary conditions are obtained for radial stresses at outside and inside surfaces by making these equal to coolant pressure and internal pressure respectively. A multi-mechanism creep model which accounts for thermal and irradiation creep is used to calculate the overall creep rate. Effective plastic strain is a function of effective stress and material constants. (orig.)

Thermal-Hydraulic Performance of Cross-Shaped Spiral Fuel in High-Power-Density BWRs

International Nuclear Information System (INIS)

Conboy, Thomas; Hejzlar, Pavel

2006-01-01

Power up-rating of existing nuclear reactors promises to be an area of great study for years to come. One of the major approaches to efficiently increasing power density is by way of advanced fuel design, and cross-shaped spiral-fuel has shown such potential in previous studies. Our work aims to model the thermal-hydraulic consequences of filling a BWR core with these spiral-shaped pins. The helically-wound pins have a cross-section resembling a 4-petaled flower. They fill an assembly in a tight bundle, their dimensions chosen carefully such that the petals of neighboring pins contact each other at their outer-most extent in a self-supporting lattice, absent of grid spacers. Potential advantages of this design raise much optimism from a thermal-hydraulic perspective. These spiral rods possess about 40% larger surface area than traditional rods, resulting in increased cooling and a proportional reduction in average surface heat flux. The thin petal-like extensions help by lowering thermal resistance between the hot central region of the pin and the bulk coolant flow, decreasing the maximum fuel temperature by 200 deg. C according to Finite Element (COSMOS) models. However, COSMOS models also predict a potential problem area at the 'elbow' region of two adjoining petals, where heat flux peaking is twice that along the extensions. Preliminary VIPRE models, which account only for the surface area increase, predict a 22% increase in critical power. It is also anticipated that the spiral twist would provide the flowing coolant with an additional radial velocity component, and likely promote turbulence and mixing within an assembly. These factors are expected to provide further margin for increased power density, and are currently being incorporated into the VIPRE model. The reduction in pressure drop inherent in any core without grid-spacers is also expected to be significant in aiding core stability, though this has not yet been quantified. Spiral-fuel seems to be a

Study on thermal performance and margins of BWR fuel elements

International Nuclear Information System (INIS)

Stosic, Zoran

1999-01-01

This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

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).

Thermally regenerative hydrogen/oxygen fuel cell power cycles

Science.gov (United States)

Morehouse, J. H.

1986-01-01

Two innovative thermodynamic power cycles are analytically examined for future engineering feasibility. The power cycles use a hydrogen-oxygen fuel cell for electrical energy production and use the thermal dissociation of water for regeneration of the hydrogen and oxygen. The TDS (thermal dissociation system) uses a thermal energy input at over 2000 K to thermally dissociate the water. The other cycle, the HTE (high temperature electrolyzer) system, dissociates the water using an electrolyzer operating at high temperature (1300 K) which receives its electrical energy from the fuel cell. The primary advantages of these cycles is that they are basically a no moving parts system, thus having the potential for long life and high reliability, and they have the potential for high thermal efficiency. Both cycles are shown to be classical heat engines with ideal efficiency close to Carnot cycle efficiency. The feasibility of constructing actual cycles is investigated by examining process irreversibilities and device efficiencies for the two types of cycles. The results show that while the processes and devices of the 2000 K TDS exceed current technology limits, the high temperature electrolyzer system appears to be a state-of-the-art technology development. The requirements for very high electrolyzer and fuel cell efficiencies are seen as determining the feasbility of the HTE system, and these high efficiency devices are currently being developed. It is concluded that a proof-of-concept HTE system experiment can and should be conducted.

Thermal behavior analysis of U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2004-07-01

According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

Thermal behavior analysis of U-Mo/Al dispersion fuel

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu

2004-01-01

According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

TRISO fuel thermal simulations in the LS-VHTR

Energy Technology Data Exchange (ETDEWEB)

Ramos, Mario C.; Scari, Maria E.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F., E-mail: marc5663@gmail.com, E-mail: melizabethscari@yahoo.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

2017-07-01

The liquid-salt-cooled very high-temperature reactor (LS-VHTR) is a reactor that presents very good characteristics in terms of energy production and safety aspects. It uses as fuel the TRISO particles immersed in a graphite matrix with a cylindrical shape called fuel compact, as moderator graphite and as coolant liquid salt Li{sub 2}BeF{sub 4} called Flibe. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core by performing different simplifications to represent the reactor core and the fuel compact under steady-state conditions, starting the modeling from a single fuel element, until complete the studies with the entire core model developed in the RELAP5-3D code. Two models were considered for representation of the fuel compact, homogeneous and non-homogeneous models, as well as different geometries of the heat structures was considered. The aim to develop several models was to compare the thermal hydraulic characteristics resulting from the construction of a more economical and less discretized model with much more refined models that can lead to more complexes analyzes to representing TRISO effect particles in the fuel compact. The different results found, mainly, for the core temperature distributions are presented and discussed. (author)

Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis

Science.gov (United States)

2016-07-31

distribution unlimited Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis Matthew...vital importance for hydrocarbon -fueled propulsion systems: fuel thermal performance as indicated by physical and chemical effects of cooling passage... analysis . The selection and acquisition of a set of chemically diverse fuels is pivotal for a successful outcome since test method validation and

Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

International Nuclear Information System (INIS)

Lightston, M.F.; Rock, R.

1996-01-01

This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

Thermal Cooling Limits of Sbotaged Spent Fuel Pools

Energy Technology Data Exchange (ETDEWEB)

Dr. Thomas G. Hughes; Dr. Thomas F. Lin

2010-09-10

To develop the understanding and predictive measures of the post “loss of water inventory” hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

Spent fuel dry storage technology development: report of consolidated thermal data

International Nuclear Information System (INIS)

Lundberg, W.L.

1980-09-01

Experiments indicate that PWR fuel with decay heat levels in excess of 2 kW could be stored in isolated drywells in Nevada Test Site soil without exceeding the current fuel clad temperature limit (715 0 F). The document also assesses the ability to thermally analyze near-surface drywells and above-ground storage casks and it identifies analysis development areas. It is concluded that the required analysis procedures, computer programs, etc., are already developed and available. Analysis uncertainties, however, still exist but they lie mainly in the numerical input area. Soil thermal conductivity, of primary importance in analysis, requires additional study to better understand the soil drying mechanism and effects of moisture. Work is also required to develop an internal canister subchannel model. In addition, the ability of the overall drywell thermal model to accommodate thermal interaction effects between adjacent drywells should be confirmed. In the experimental area, tests with two BWR spent fuel assemblies encapsulated in a single canister should be performed to establish the fuel clad and canister temperature relationship. This is needed to supplement similar experimental work which has already been completed with PWR fuel

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Thermal signature measurements for ammonium nitrate/fuel mixtures by laser heating

International Nuclear Information System (INIS)

Nazarian, Ashot; Presser, Cary

2016-01-01

Highlights: • LDTR is a useful diagnostic for characterizing AN/fuel mixture thermochemical behavior. • Each AN/fuel mixture thermal signature was different. • AN/fuel mixture signature features were defined by the individual constituents. • Baseline signatures changed after an experiment. - Abstract: Measurements were carried out to obtain thermal signatures of several ammonium nitrate/fuel (ANF) mixtures, using a laser-heating technique referred to as the laser-driven thermal reactor (LDTR). The mixtures were ammonium nitrate (AN)/kerosene, AN/ethylene glycol, AN/paraffin wax, AN/petroleum jelly, AN/confectioner's sugar, AN/cellulose (tissue paper), nitromethane/cellulose, nitrobenzene/cellulose, AN/cellulose/nitromethane, AN/cellulose/nitrobenzene. These mixtures were also compared with AN/nitromethane and AN/diesel fuel oil, obtained from an earlier investigation. Thermograms for the mixtures, as well as individual constituents, were compared to better understand how sample thermal signature changes with mixture composition. This is the first step in development of a thermal-signature database, to be used along with other signature databases, to improve identification of energetic substances of unknown composition. The results indicated that each individual thermal signature was associated unambiguously with a particular mixture composition. The signature features of a particular mixture were shaped by the individual constituent signatures. It was also uncovered that the baseline signature was modified after an experiment due to coating of unreacted residue on the substrate surface and a change in the reactor sphere oxide layer. Thus, care was required to pre-oxidize the sphere prior to an experiment. A minimum sample mass (which was dependent on composition) was required to detect the signature characteristics. Increased laser power served to magnify signal strength while preserving the signature features. For the mixtures examined, the thermal

Thermal conductivity thermal diffusivity of UO{sub 2}-BeO nuclear fuel pellets

Energy Technology Data Exchange (ETDEWEB)

Mansur, Fábio A.; Camarano, Denise M.; Santos, Ana M. M.; Ferraz, Wilmar B.; Silva, Mayra A.; Ferreira, Ricardo A.N., E-mail: fam@cdtn.br, E-mail: dmc@cdtn.br, E-mail: amms@cdtn.br, E-mail: ferrazw@cdtn.br, E-mail: mayra.silva@cdtn.br, E-mail: ricardoanf@yahoo.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The temperature distribution in nuclear fuel pellets is of vital importance for the performance of the reactor, as it affects the heat transfer, the mechanical behavior and the release of fission gas during irradiation, reducing safety margins in possible accident scenarios. One of the main limitation for the current uranium dioxide nuclear fuel (UO{sub 2}) is its low thermal conductivity, responsible for the higher temperature of the pellet center and, consequently, for a higher radial temperature gradient. Thus, the addition of another material to increase the UO{sub 2} fuel thermal conductivity has been considered. Among the additives that are being investigated, beryllium oxide (BeO) has been chosen due to its high thermal conductivity, with potential to optimize power generation in pressurized light water reactors (PWR). In this work, UO{sub 2}-BeO pellets were obtained by the physical mixing of the powders with additions of 2wt% and 3wt% of BeO. The thermal diffusivity and conductivity of the pellets were determined from room temperature up to 500 °C. The results were normalized to 95% of the theoretical density (TD) of the pellets and varied according to the BeO content. The range of the values of thermal diffusivity and conductivity were 1.22 mm{sup 2}∙s{sup -1} to 3.69 mm{sup 2}∙s{sup -1} and 3.80 W∙m{sup -}'1∙K{sup -1} to 9.36 W∙m{sup -1}∙K{sup -1}, respectively. (author)

Online ICPMS detection of the thermal release of fission products from nuclear fuel samples

International Nuclear Information System (INIS)

Guenther-Leopold, I.; Svedkauskaite-Le Gore, J.; Kivel, N.

2009-01-01

Full text: The release of volatile and semi-volatile fission products (like Cs, Tc, Mo etc.) from spent nuclear fuel by thermal and thermochemical treatment (oxidative or reductive conditions) as a head-end step for advanced reprocessing scenarios is studied in the Hot Laboratory of the Paul Scherrer Institut. For this purpose, a heated sampling cell online connected to an ICPMS (Element 2, Thermo Fisher Scientific) was designed and tested on simulated fuel samples up to 650 o C. The results of this study as well as technical perspectives for heating experiments up to 2000 o C will be presented. (author)

Thermal expansion of UO2-Gd2O3 fuel pellets

International Nuclear Information System (INIS)

Une, Katsumi

1986-01-01

In recent years, more consideration has been given to the application of UO 2 -Gd 2 O 3 burnable poison fuel to LWRs in order to improve the core physics and to extend the burnup. It has been known that UO 2 forms a single phase cubic fluorite type solid solution with Gd 2 O 3 up to 20 - 30 wt.% above 1300 K. The addition of Gd 2 O 3 to UO 2 lattices changes the properties of the fuel pellets. The limited data on the thermal expansion of UO 2 -Gd 2 O 3 fuel exist, but those are inconsistent. UO 2 -Gd 2 O 3 fuel pellets were fabricated, and the linear thermal expansion of UO 2 and UO 2 -(5, 8 and 10 wt.%)Gd 2 O 3 fuel pellets was measured with a differential dilatometer over the temperature range of 298 - 1973 K. A sapphire rod of 6 mm diameter and 15.5 mm length was used as the reference material. After the preheating cycle, the measurement was performed in argon atmosphere. The results for UO 2 pellets showed excellent agreement with the data in literatures. The linear thermal expansion of UO 2 -Gd 2 O 3 fuel pellets showed the increase with increasing the Gd 2 O 3 content. Consideration must be given to this excessive expansion in the fuel design of UO 2 -Gd 2 O 3 pellets. The equations for the linear thermal expansion and density of UO 2 -Gd 2 O 3 fuel pellets were derived by the method of least squares. (Kako, I.)

Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

International Nuclear Information System (INIS)

Gwaltney, R.C.; McAfee, W.J.

1977-01-01

A new method for performing thermal stress analyses in structures with multiple penetrations was applied to these analyses. This method couples the development of an equivalent thermal conductivity for the blocks, a technique that has been used extensively for modeling the thermal characteristics of reactor cores, with the use of the equivalent solid plate method for stress analysis. Using this equivalent thermal conductivity, which models as one material the heat transfer characteristics of the fuel, coolant, and graphite two-dimensional, steady-state thermal analyses of the fuel and control rod fuel blocks were performed to establish all temperature boundaries required for the stress analyses. In applying the equivalent solid plate method, the region of penetrations being modeled was replaced by a pseudo material having the same dimensions but whose materials properties were adjusted to account for the penetration. The peak stresses and strains were determined by applying stress and strain intensification factors to the calculated distributions. The condition studied was where the blocks were located near the center of the furnace. In this position, the axial surface of the block is heated near one end and cooled near the other. The approximate axial surface temperatures ranged from 1521 0 C at both the heated and the cooled ends to a peak of 1800 0 C near the center. Five specific cases were analyzed: plane (two-dimensional thermal, plane stress strain) analyses of each end of a standard fuel block (2 cases), plane analyses of each end of a control rod fuel block (2 cases), and a two-dimensional analysis of a fuel block treated as an axisymmetric cylind

Experimental determination of thermal conductivity and gap conductance of fuel rod for HTGR

International Nuclear Information System (INIS)

Kikuchi, Teruo; Iwamoto, Kazumi; Ikawa, Katsuichi; Ishimoto, Kiyoshi

1985-01-01

The thermal conductivity of fuel compacts and the gap conductance between the fuel compact and the graphite sleeve in fuel rods for a high-temperature gas-cooled reactor (HTGR) were measured by the center heating method. These measurements were made as functions of volume percent particle loading and temperature for thermal conductivity and as functions of gap distance and gas composition for gap conductance. The thermal conductivity of fuel compacts decreases with increasing temperature and with increasing particle loading. The gap conductance increases with increasing temperature and decrease with increasing gap distance. A good gap conductance was observed with helium fill gas. It was seen that the gap conductance was dependent on the thermal conductivity of fill gas and conductance by radiation and could be neglected the conductance through solid-solid contact points of fuel compact and graphite sleeve. (author)

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

Mathematical model of thermal and mechanical steady state fuel element behaviour TEDEF

International Nuclear Information System (INIS)

Dinic, N.; Kostic, Z.; Josipovic, M.

1987-01-01

In this paper a numerical model of thermal and thermomechanical behaviour of a cylindrical metal uranium fuel element is described. Presented are numerical method and computer program for solving the stationary temperature field and thermal stresses of a fuel element. The model development is a second phase of analysis of these phenomena, and may as well be used for analysing power nuclear reactor fuel elements behaviour. (author)

Uncertainty and sensitivity analysis of the nuclear fuel thermal behavior

Energy Technology Data Exchange (ETDEWEB)

Boulore, A., E-mail: antoine.boulore@cea.fr [Commissariat a l' Energie Atomique (CEA), DEN, Fuel Research Department, 13108 Saint-Paul-lez-Durance (France); Struzik, C. [Commissariat a l' Energie Atomique (CEA), DEN, Fuel Research Department, 13108 Saint-Paul-lez-Durance (France); Gaudier, F. [Commissariat a l' Energie Atomique (CEA), DEN, Systems and Structure Modeling Department, 91191 Gif-sur-Yvette (France)

2012-12-15

Highlights: Black-Right-Pointing-Pointer A complete quantitative method for uncertainty propagation and sensitivity analysis is applied. Black-Right-Pointing-Pointer The thermal conductivity of UO{sub 2} is modeled as a random variable. Black-Right-Pointing-Pointer The first source of uncertainty is the linear heat rate. Black-Right-Pointing-Pointer The second source of uncertainty is the thermal conductivity of the fuel. - Abstract: In the global framework of nuclear fuel behavior simulation, the response of the models describing the physical phenomena occurring during the irradiation in reactor is mainly conditioned by the confidence in the calculated temperature of the fuel. Amongst all parameters influencing the temperature calculation in our fuel rod simulation code (METEOR V2), several sources of uncertainty have been identified as being the most sensitive: thermal conductivity of UO{sub 2}, radial distribution of power in the fuel pellet, local linear heat rate in the fuel rod, geometry of the pellet and thermal transfer in the gap. Expert judgment and inverse methods have been used to model the uncertainty of these parameters using theoretical distributions and correlation matrices. Propagation of these uncertainties in the METEOR V2 code using the URANIE framework and a Monte-Carlo technique has been performed in different experimental irradiations of UO{sub 2} fuel. At every time step of the simulated experiments, we get a temperature statistical distribution which results from the initial distributions of the uncertain parameters. We then can estimate confidence intervals of the calculated temperature. In order to quantify the sensitivity of the calculated temperature to each of the uncertain input parameters and data, we have also performed a sensitivity analysis using the Sobol' indices at first order.

Using plasma-fuel systems at Eurasian coal-fired thermal power stations

Science.gov (United States)

Karpenko, E. I.; Karpenko, Yu. E.; Messerle, V. E.; Ustimenko, A. B.

2009-06-01

The development of plasma technology for igniting solid fuels at coal-fired thermal power stations in Russia, Kazakhstan, China, and other Eurasian countries is briefly reviewed. Basic layouts and technical and economic characteristics of plasma-fuel systems installed in different coal-fired boiles are considered together with some results from using these systems at coal-fired thermal power stations.

Ground test facilities for evaluating nuclear thermal propulsion engines and fuel elements

International Nuclear Information System (INIS)

Allen, G.C.; Beck, D.F.; Harmon, C.D.; Shipers, L.R.

1992-01-01

Interagency panels evaluating nuclear thermal propulsion development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and design issues of a proposed ground test complex for evaluating nuclear thermal propulsion engines and fuel elements being developed for the Space Nuclear Thermal Propulsion (SNTP) program. 2 refs

Effect of long-term storage of LWR spent fuel on Pu-thermal fuel cycle

International Nuclear Information System (INIS)

Kurosawa, Masayoshi; Naito, Yoshitaka; Suyama, Kenya; Itahara, Kuniyuki; Suzuki, Katsuo; Hamada, Koji

1998-01-01

According to the Long-term Program for Research, Development and Utilization of Nuclear Energy (June, 1994) in Japan, the Rokkasho Reprocessing Plant will be operated shortly after the year 2000, and the planning of the construction of the second commercial plant will be decided around 2010. Also, it is described that spent fuel storage has a positive meaning as an energy resource for the future utilization of Pu. Considering the balance between the increase of spent fuels and the domestic reprocessing capacity in Japan, it can be expected that the long-term storage of UO 2 spent fuels will be required. Then, we studied the effect of long-term storage of spent fuels on Pu-thermal fuel cycle. The burnup calculation were performed on the typical Japanese PWR fuel, and the burnup and criticality calculations were carried out on the Pu-thermal cores with MOX fuel. Based on the results, we evaluate the influence of extending the spent fuel storage term on the criticality safety, shielding design of the reprocessing plant and the core life time of the MOX core, etc. As the result of this work on long-term storage of LWR spent fuels, it becomes clear that there are few demerits regarding the lifetime of a MOX reactor core, and that there are many merits regarding the safety aspects of the fuel cycle facilities. Furthermore, long-term storage is meaningful as energy storage for effective utilization of Pu to be improved by technological innovation in future, and it will allow for sufficient time for the important policymaking of nuclear fuel cycle establishment in Japan. (author)

Modification of fuel properties under thermal load; Veraenderung von Kraftstoffeigenschaften unter thermischer Belastung

Energy Technology Data Exchange (ETDEWEB)

Lau, Kornelia; Richter, Beate; Schuemann, Ulrike; Crusius, Svetlana; Streibel, Thorsten W.; Harndorf, Horst [Rostock Univ. (Germany). Abt. Analytische und Technische Chemie

2013-10-01

Diesel fuel without FAME and additives as well as rapeseed methyl ester were thermally stressed (150 C) at laboratory conditions. In the course of the performed study, the chemical composition of the fuels and possibly generated residues were analyzed with regards to the influence of oxygen availability and test duration. Therefore, chromatographic methods as GC-MS (Gas Chromatography-Mass Spectrometry) and HP-SEC (High Performance-Size Exclusion Chromatography) have successfully been applied. In the formed solid B 0-residues mainly oxygen-containing aromatic compounds were identified, while the biodiesel samples show mostly decomposition products of the biofuel like carbonic acids and aldehydes. In both fuels an increasing amount of compounds with high molecular weights was observed versus test duration. It is assumed that these components consist of oligo- and polymers with high oxygen content. However, this needs to be proved by further analytical methods. On the basis of these results prospective tailor-made additive packages for the stabilisation of fuels and the prevention of deposits in fuel injection systems, can be developed and applied. (orig.)

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

International Nuclear Information System (INIS)

Kosaka, Yuji; Kurematsu, Shigeru; Kitagawa, Takaaki; Suzuki, Akihiro; Terai, Takayuki

2012-01-01

The temperature measurements of mixed oxide (MOX) and UO 2 fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO 2 fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel. (author)

Cellular Scanning Strategy for Selective Laser Melting: Capturing Thermal Trends with a Low-Fidelity, Pseudo-Analytical Model

Directory of Open Access Journals (Sweden)

Sankhya Mohanty

2014-01-01

Full Text Available Simulations of additive manufacturing processes are known to be computationally expensive. The resulting large runtimes prohibit their application in secondary analysis requiring several complete simulations such as optimization studies, and sensitivity analysis. In this paper, a low-fidelity pseudo-analytical model has been introduced to enable such secondary analysis. The model has been able to mimic a finite element model and was able to capture the thermal trends associated with the process. The model has been validated and subsequently applied in a small optimization case study. The pseudo-analytical modelling technique is established as a fast tool for primary modelling investigations.

Stress relaxation of thermally bowed fuel pins

International Nuclear Information System (INIS)

Crossland, I.G.; Speight, M.V.

1983-01-01

The presence of cross-pin temperature gradients in nuclear reactor fuel pins produces differential thermal expansion which, in turn, causes the fuel pin to bow elastically. If the pin is restrained in any way, such thermal bowing causes the pin to be stressed. At high temperatures these stresses can relax by creep and it is shown here that this causes the pin to suffer an additional permanent deflection, so that when the cross-pin temperature difference is removed the pin remains bowed. By representing the cylindrical pin by an equivalent I-beam, the present work examines this effect when it takes place by secondary creep. Two restraint systems are considered, and it is demonstrated that the rate of relaxation depends mainly upon the creep equation, and hence the temperature, and also the magnitude of the initial stresses. (author)

Spent nuclear fuel storage pool thermal-hydraulic analysis

International Nuclear Information System (INIS)

Gay, R.R.

1984-01-01

Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

Assessment of effective thermal conductivity in U–Mo metallic fuels with distributed gas bubbles

Energy Technology Data Exchange (ETDEWEB)

Hu, Shenyang; Casella, Andrew M.; Lavender, Curt A.; Senor, David J.; Burkes, Douglas E.

2015-07-15

This work presents a numerical method to assess the relative impact of various microstructural features including grain sizes, nanometer scale intragranular gas bubbles, and larger intergranular gas bubbles in irradiated U–Mo metallic fuels on the effective thermal conductivity. A phase-field model was employed to construct a three-dimensional polycrystalline U–Mo fuel alloy with a given crystal morphology and gas bubble microstructures. An effective thermal conductivity “concept” was taken to capture the effect of polycrystalline structures and gas bubble microstructures with significant size differences on the thermal conductivity. The thermal conductivity of inhomogeneous materials was calculated by solving the heat transport equation. The obtained results are in reasonably good agreement with experimental measurements made on irradiated U–Mo fuel samples containing similar microstructural features. The developed method can be used to predict the thermal conductivity degradation in operating nuclear fuels if the evolution of microstructures is known during operation of the fuel.

Thermal-hydraulic effects of transition to improved System 80TM fuel

International Nuclear Information System (INIS)

Rodack, T.; Joffre, P.F.; Kapoor, R.K.

2004-01-01

ABB CE's improved System 80 TM PWR fuel design includes GUARDIAN debris-resistant features and laser-welded Zircaloy grids. The GUARDIAN features include an Inconel grid with debris-filtering features located just above the Lower End Fitting, and a solid fuel rod bottom end cap that extends above the filtering features. Tests and analyses were done to establish the impact of these design improvements on fuel assembly hydraulic performance. Further analysis was done to determine the mixed core thermal-hydraulic performance as the transition is made over two fuel cycles to a full core of the improved System 80 TM fuel. Results confirm that the Thermal-Hydraulic (T-H) effects of the reduction in hydraulic resistance between the improved and resident fuel due to the laser-welded Zircaloy grids offsets the effects of the increased resistance GUARDIAN grid. Therefore, the mechanically improved System 80 TM fuel can be implemented with no net impact on Departure from Nucleate Boiling (DNB) margin in transition cores. (author)

High Fidelity Simulation of Primary Atomization in Diesel Engine Sprays

Science.gov (United States)

Ivey, Christopher; Bravo, Luis; Kim, Dokyun

2014-11-01

A high-fidelity numerical simulation of jet breakup and spray formation from a complex diesel fuel injector at ambient conditions has been performed. A full understanding of the primary atomization process in fuel injection of diesel has not been achieved for several reasons including the difficulties accessing the optically dense region. Due to the recent advances in numerical methods and computing resources, high fidelity simulations of atomizing flows are becoming available to provide new insights of the process. In the present study, an unstructured un-split Volume-of-Fluid (VoF) method coupled to a stochastic Lagrangian spray model is employed to simulate the atomization process. A common rail fuel injector is simulated by using a nozzle geometry available through the Engine Combustion Network. The working conditions correspond to a single orifice (90 μm) JP-8 fueled injector operating at an injection pressure of 90 bar, ambient condition at 29 bar, 300 K filled with 100% nitrogen with Rel = 16,071, Wel = 75,334 setting the spray in the full atomization mode. The experimental dataset from Army Research Lab is used for validation in terms of spray global parameters and local droplet distributions. The quantitative comparison will be presented and discussed. Supported by Oak Ridge Associated Universities and the Army Research Laboratory.

Thermal hydraulic model validation for HOR mixed core fuel management

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

1997-01-01

A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

An assessment of thermal behavior of the DUPIC fuel bundle by subchannel analysis

International Nuclear Information System (INIS)

Park, Jee Won.

1997-12-01

Thermal behavior of the standard DUPIC fuel has been assessed. The DUPIC fuel bundle has been modeled for a subchannel analysis using the ASSERT-IV code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions of the DUPIC fuel bundle, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. Based upon the subchannel modeling used in this study, the location of minimum CHFR in the DUPIC fuel bundle has been found to be very similar to that of the standard fuel. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction was found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. Since the transverse interchange model between subchannels is important for the behavior of these variables, it is needed to put more effort in validating the transverse interchange model. For the purpose of investigating influence of thermal-hydraulic parameter variations of the DUPIC fuel bundle, four different values of the channel flow rates were used in the subchannel analysis. The effect of the channel flow reduction on thermal-hydraulic parameters have been presented. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundles in CANDU reactors. (author). 12 refs., 3 tabs., 17 figs

Thermal modeling of the ceramic composite fuel for light water reactors

International Nuclear Information System (INIS)

Revankar, S.T.; Latta, R.; Solomon, A.A.

2005-01-01

Full text of publication follows: Composite fuel designs capable of providing improved thermal performance are of great interest in advanced reactor designs where high efficiency and long fuel cycles are desired. Thermal modeling of the composite fuel consisting of continuous second phase in a ceramic (uranium oxide) matrix has been carried out with detailed examination of the microstructure of the composite and the interface. Assuming that constituent phases are arranged as slabs, upper and lower bounds for the thermal conductivity of the composite are derived analytically. Bounding calculations on the thermal conductivity of the composite were performed for SiC dispersed in the UO 2 matrix. It is found that with 10% SiC, the thermal conductivity increases from 5.8 to 9.8 W/m.deg. K at 500 K, or an increase of 69% was observed in UO 2 matrix. The finite element analysis computer program ANSYS was used to create composite fuel geometries with set boundary conditions to produce accurate thermal conductivity predictions. A model developed also accounts for SiC-matrix interface resistance and the addition of coatings or interaction barriers. The first set of calculations using the code was to model simple series and parallel fuel slab geometries, and then advance to inter-connected parallel pathways. The analytical calculations were compared with the ANSYS results. The geometry of the model was set up as a 1 cm long by 400 micron wide rectangle. This rectangle was then divided into one hundred sections with the first ninety percent of a single section being UO 2 and the remaining ten percent consisting of SiC. The model was then meshed using triangular type elements. The boundary conditions were set with the sides of the rectangle being adiabatic and having an assigned temperature at the end of the rectangle. A heat flux was then applied to one end of the model producing a temperature gradient. The effective thermal conductivity was then calculated using the geometry

Thermal degradation of ligno-cellulosic fuels. DSC and TGA studies

Energy Technology Data Exchange (ETDEWEB)

Leroy, V.; Cancellieri, D.; Leoni, E. [SPE-CNRS UMR 6134, University of Corsica, Campus Grossetti, BP 52, 20250 Corti (France)

2006-12-01

The scope of this work was to show the utility of thermal analysis and calorimetric experiments to study the thermal oxidative degradation of Mediterranean scrubs. We investigated the thermal degradation of four species; DSC and TGA were used under air sweeping to record oxidative reactions in dynamic conditions. Heat released and mass loss are important data to be measured for wildland fires modelling purpose and fire hazard studies on ligno-cellulosic fuels. Around 638 and 778K, two dominating and overlapped exothermic peaks were recorded in DSC and individualized using a experimental and numerical separation. This stage allowed obtaining the enthalpy variation of each exothermic phenomenon. As an application, we propose to classify the fuels according to the heat released and the rate constant of each reaction. TGA experiments showed under air two successive mass loss around 638 and 778K. Both techniques are useful in order to measure ignitability, combustibility and sustainability of forest fuels. (author)

Analytical determination of thermal conductivity of W-UO2 and W-UN CERMET nuclear fuels

Science.gov (United States)

Webb, Jonathan A.; Charit, Indrajit

2012-08-01

The thermal conductivity of tungsten based CERMET fuels containing UO2 and UN fuel particles are determined as a function of particle geometry, stabilizer fraction and fuel-volume fraction, by using a combination of an analytical approach and experimental data collected from literature. Thermal conductivity is estimated using the Bruggeman-Fricke model. This study demonstrates that thermal conductivities of various CERMET fuels can be analytically predicted to values that are very close to the experimentally determined ones.

Evaluating display fidelity and interaction fidelity in a virtual reality game.

Science.gov (United States)

McMahan, Ryan P; Bowman, Doug A; Zielinski, David J; Brady, Rachael B

2012-04-01

In recent years, consumers have witnessed a technological revolution that has delivered more-realistic experiences in their own homes through high-definition, stereoscopic televisions and natural, gesture-based video game consoles. Although these experiences are more realistic, offering higher levels of fidelity, it is not clear how the increased display and interaction aspects of fidelity impact the user experience. Since immersive virtual reality (VR) allows us to achieve very high levels of fidelity, we designed and conducted a study that used a six-sided CAVE to evaluate display fidelity and interaction fidelity independently, at extremely high and low levels, for a VR first-person shooter (FPS) game. Our goal was to gain a better understanding of the effects of fidelity on the user in a complex, performance-intensive context. The results of our study indicate that both display and interaction fidelity significantly affect strategy and performance, as well as subjective judgments of presence, engagement, and usability. In particular, performance results were strongly in favor of two conditions: low-display, low-interaction fidelity (representative of traditional FPS games) and high-display, high-interaction fidelity (similar to the real world).

DART model for thermal conductivity of U3Si2 aluminum dispersion fuel

International Nuclear Information System (INIS)

Rest, J.; Snelgrove, J.L.; Hofman, G.L.

1995-09-01

This paper describes the primary physical models that form the basis of the DART model for calculating irradiation-induced changes in the thermal conductivity of aluminium dispersion fuel. DART calculations of fuel swelling, pore closure, and thermal conductivity are compared with measured values

Resin bead-thermal ionization mass spectrometry for determination of plutonium concentration in irradiated fuel dissolver solution

International Nuclear Information System (INIS)

Paul, Sumana; Shah, R.V.; Aggarwal, S.K.; Pandey, A.K.

2015-01-01

Determination of isotopic composition (IC) and concentration of plutonium (Pu) is necessary at various stages of nuclear fuel cycle which involves analysis of complex matrices like dissolver solution of irradiated fuel, nuclear waste stream etc. Mass spectrometry, e.g. thermal ionization mass spectrometry (TIMS) and inductively coupled plasma mass spectrometry (ICP-MS) are commonly used for determination of IC and concentration of plutonium. However, to circumvent matrix interferences, efficient separation as well as preconcentration of Pu is required prior to mass spectrometric analysis. Purification steps employing ion-exchange resins are widely used for the separation of Pu from dissolver solution or from mixture of other actinides e.g. U, Am. However, an alternative way is to selectively preconcentrate Pu on a resin bead, followed by direct loading of the bead on the filament of thermal ionization mass spectrometer

Thermal barrier and support for nuclear reactor fuel core

International Nuclear Information System (INIS)

Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

1987-01-01

A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

Thermal stress analysis of HTGR fuel and control rod fuel blocks in the HTGR in-block carbonization and annealing furnace

International Nuclear Information System (INIS)

Gwaltney, R.C.; McAfee, W.J.

1977-01-01

A new approach that utilizes the equivalent solid plate method has been applied to the thermal stress analysis of HTGR fuel and control rod fuel blocks. Cases were considered where these blocks, loaded with reprocessed HTGR fuel pellets, were being cured at temperatures up to 1800 0 C. A two-dimensional segment of a fuel block cross section including fuel, coolant holes, and graphite matrix was analyzed using the ORNL HEATING3 heat transfer code to determine the temperature-dependent effective thermal conductivity for the perforated region of the block. Using this equivalent conductivity to calculate the temperature distributions through different cross sections of the blocks, two-dimensional thermal-stress analyses were performed through application of the equivalent solid plate method. In this approach, the perforated material is replaced by solid homogeneous material of the same external dimensions but whose material properties have been modified to account for the perforations

Waste heat recovery for transport trucks using thermally regenerative fuel cells

Energy Technology Data Exchange (ETDEWEB)

Carrier, A.; Wechsler, D.; Whitney, R.; Jessop, P. [Queen' s Univ., Kingston, ON (Canada). Dept. of Chemistry; Davis, B.R. [Queen' s-RMC Fuel Cell Research Centre, Kingston, ON (Canada)

2009-07-01

Carbon emissions associated with transportation can be reduced by increasing the fuel efficiency of transport trucks. This can be achieved with thermally regenerative fuel cells that transform the waste heat from the engine block into electricity. In order to operate such a fuel cell, one needs a fluid which rapidly, reversibly, and selectively undergoes dehydrogenation. Potential fluids have been screened for their ability to dehydrogenate and then rehydrogenate at the appropriate temperatures. An examination of the thermodynamics, kinetics, and selectivities of these processes have shown that the challenge involving hydrogenolysis at high temperature must be addressed. This paper discussed the economics of thermally regenerative fuel cells and the advantages and disadvantages of the identified fluids, and of such systems in general.

PEM fuel cells thermal and water management fundamentals

CERN Document Server

Wang, Yun; Cho, Sung Chan

2014-01-01

Polymer Electrolyte Membrane (PEM) fuel cells convert chemical energy in hydrogen into electrical energy with water as the only by-product. Thus, PEM fuel cells hold great promise to reduce both pollutant emissions and dependency on fossil fuels, especially for transportation-passenger cars, utility vehicles, and buses-and small-scale stationary and portable power generators. But one of the greatest challenges to realizing the high efficiency and zero emissions potential of PEM fuel cells technology is heat and water management. This book provides an introduction to the essential concepts for effective thermal and water management in PEM fuel cells and an assessment on the current status of fundamental research in this field. The book offers you: An overview of current energy and environmental challenges and their imperatives for the development of renewable energy resources, including discussion of the role of PEM fuel cells in addressing these issues; Reviews of basic principles pertaining to PEM fuel cel...

DART model for thermal conductivity of U3Si2 Aluminum dispersion fuel

International Nuclear Information System (INIS)

Rest, J.; Snelgrove, J.L.; Hofman, G.L.

2004-01-01

This paper describes the primary physical models that form the basis of the DART model for calculating irradiation-induced changes in the thermal conductivity of aluminum dispersion fuel. DART calculations of fuel swelling, pore closure, and thermal conductivity are compared with measured values. (author)

Thermal Effects by Firing Oil Shale Fuel in CFB Boilers

Science.gov (United States)

Neshumayev, D.; Ots, A.; Parve, T.; Pihu, T.; Plamus, K.; Prikk, A.

It is well known that during firing of oil shale fuel the amount of heat released during its combustion per kg of fuel is significantly affected by the endothermic and exothermic processes taking place in mineral matter. These thermal effects are calcite and dolomite decomposing, marcasite FeS2 oxidising, CaO sulphation and formation of the new minerals. The given paper deals with the experimental study of the influence of these thermal effects of oil shale fuel having different heating value on total amount of heat released during combustion in calorimetric bomb, circulating fluidized bed (CFB) and pulverized-firing boiler (PFB). The large-scale (250 MWth) experiments were performed in the K11-1 CFB boiler of the Balti Power Plant. During experiments low heating value of a fuel varied within the range 8.5-11 MJ/kg. At the end some conclusions were drawn.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

Energy Technology Data Exchange (ETDEWEB)

Mo, Kun, E-mail: kunmo@anl.gov; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

2017-04-15

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO{sub 2} particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO{sub 2} particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO{sub 2} particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO{sub 2} particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Experimental Measurement and Numerical Modeling of the Effective Thermal Conductivity of TRISO Fuel Compacts

International Nuclear Information System (INIS)

Folsom, Charles

2015-01-01

Accurate modeling capability of thermal conductivity of tristructural-isotropic (TRISO) fuel compacts is important to fuel performance modeling and safety of Generation IV reactors. To date, the effective thermal conductivity (ETC) of tristructural-isotropic (TRISO) fuel compacts has not been measured directly. The composite fuel is a complicated structure comprised of layered particles in a graphite matrix. In this work, finite element modeling is used to validate an analytic ETC model for application to the composite fuel material for particle-volume fractions up to 40%. The effect of each individual layer of a TRISO particle is analyzed showing that the overall ETC of the compact is most sensitive to the outer layer constituent. In conjunction with the modeling results, the thermal conductivity of matrix-graphite compacts and the ETC of surrogate TRISO fuel compacts have been successfully measured using a previously developed measurement system. The ETC of the surrogate fuel compacts varies between 50-30 W m -1 K -1 over a temperature range of 50-600°C. As a result of the numerical modeling and experimental measurements of the fuel compacts, a new model and approach for analyzing the effect of compact constituent materials on ETC is proposed that can estimate the fuel compact ETC with approximately 15-20% more accuracy than the old method. Using the ETC model with measured thermal conductivity of the graphite matrix-only material indicate that, in the composite form, the matrix material has a much greater thermal conductivity, which is attributed to the high anisotropy of graphite thermal conductivity. Therefore, simpler measurements of individual TRISO compact constituents combined with an analytic ETC model, will not provide accurate predictions of overall ETC of the compacts emphasizing the need for measurements of composite, surrogate compacts.

Conventional OTSG development for heavy liquid fuel firing in thermal applications

Energy Technology Data Exchange (ETDEWEB)

Setchfield, W.P. [Mitchell Engineers Ltd., Glasgow, Scotland (United Kingdom); Roset, J.N. [Total S.A., Paris (France); Schaffer, M. [Total E and P Canada Ltd., Calgary, AB (Canada); O' Connor, D. [MEG Energy Inc., Calgary, AB (Canada); Kense, K. [TIW Western Inc., Calgary, AB (Canada)

2008-10-15

The demand for natural gas is expected to increase as a result of future expansion in Canadian extra heavy oil in-situ thermal production, such as steam assisted gravity drainage or SAGD projects. Natural gas is the current predominant fuel utilized for the associated steam generation. Potential natural gas shortages and related price volatility require that operators consider alternative fuels for the projected growth of in-situ thermal production in Alberta. This paper targeted the use of bitumen from upstream sites and derivative residues from upgrading activities as the most convenient alternative fuel sources for thermal operators of established horizontal type gas fired once through steam generators (OTSGs). The paper presented the methodology, the issues associated with bitumen or residue burning and the related technical solutions in developing a multi-fuel OTSG product. The paper provided background information on conventional OTSG design development, conventional OTSG existing deign, and general description of conventional OTSG. The paper also described the configuration of a radiant furnace, convection module, and theories and definitions such as heavy liquid fuels. A description and application of the equipment and processes as well as a presentation of the data and results was then offered. The multi fuel OTSG design is considered to be a practical and workable product capable of firing heavy liquid fuels. However, the design changes have had a significant impact when compared with conventional OTSG boilers. 11 figs.

Thermal efficiency and particulate pollution estimation of four biomass fuels grown on wasteland

Energy Technology Data Exchange (ETDEWEB)

Kandpal, J.B.; Madan, M. [Indian Inst. of Tech., New Delhi (India). Centre for Rural Development and Technology

1996-10-01

The thermal performance and concentration of suspended particulate matter were studied for 1-hour combustion of four biomass fuels, namely Acacia nilotica, Leucaena leucocepholea, Jatropha curcus, and Morus alba grown in wasteland. Among the four biomass fuels, the highest thermal efficiency was achieved with Acacia nilotica. The suspended particulate matter concentration for 1-hour combustion of four biomass fuels ranged between 850 and 2,360 {micro}g/m{sup 3}.

Correlations between fuel pins irradiated in fast and thermal fluxes using the frump fuel pin modelling program

International Nuclear Information System (INIS)

Hayns, M.R.; Adam, J.

1975-08-01

There is no experimental facilities in which a fuel pin can be irradiated in a fast environment under well defined conditions of over power or flow run down. Consequently most of the infor mation which is being accumulated on the behaviour of fuel pins under severe conditions is obtained from either capsule or loop rigs in thermal reactors. It is the purpose of this paper to highlight the differences between the behaviour of fuel pins irradiated in a thermal flux and a fast flux. A typical set of conditions is taken from an overpower experiment in a thermal flux and the behaviour of the system is analysed using the fuel modelling program FRUMP. A second numerical experiment is then performed in which the same conditions prevail, except that a fast flux is assumed, the criterion for comparison being that the total power input to the system is the same in both cases. From the many possible correlations which result from such an exercise the fuel tempreature has been selected to highlight various important features of the two irradiations. It is demonstrated that the flux depression can cause differences in the pin behaviour, even to altering the order of events in a transient. For example fuel melting will occur at different times and at different positions in the fuel in the two cases. It is concluded that the techniques of fuel modelling, as typified in the program FRUMP can provide a very useful tool indeed for the analysis of such experiments and for guiding the establishment of the appropriate correlations for the extrapolation to the fast flux case. (author)

Calculated and experimental substantiation of the thermal method for non-destructive testing of fuel elements

International Nuclear Information System (INIS)

Maksimov, N.M.; Soldatenko, V.A.; Petrovichev, V.I.; Salimov, S.E.; Aleksandrov, K.A.; Kurov, D.A.

1985-01-01

The main systems and methods of thermal testing, their potentialities and advantages, thermal irradiation photodetectors are described. Possible fields of application of thermal testing in nuclear engineering are discussed. Calculations of the fuel element nonstationary temperature field in the three-dimensional geometry in the presence of such an effect as fuel exfaliation from cladding are presented. The developed method and equipment for fuel element thermal testing are described. Preliminary experimental data being in agreement with the calculated ones and opening the prospects for flaw detecting are presened

Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels.

Science.gov (United States)

Kucharski, Timothy J; Ferralis, Nicola; Kolpak, Alexie M; Zheng, Jennie O; Nocera, Daniel G; Grossman, Jeffrey C

2014-05-01

Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol(-1), and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

Development of fuel performance and thermal hydraulic technology

International Nuclear Information System (INIS)

Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

2000-03-01

Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

Fuel Assemblies Thermal Analysis in the New Spent Fuel Storage Facility at Inshass Site

International Nuclear Information System (INIS)

Khattab, M.; Mariy, Ahmed

1999-01-01

New Wet Storage Facility (NSF) is constructed at Inshass site to solve the problem of spent fuel storage capacity of ETRR-1 reactor . The Engineering Safety Heat Transfer Features t hat characterize the new facility are presented. Thermal analysis including different scenarios of pool heat load and safety limits are discussed . Cladding temperature limit during handling and storage process are specified for safe transfer of fuel

Biodiesel from vegetable oil as alternate fuel for C.I engine and feasibility study of thermal cracking: A critical review

International Nuclear Information System (INIS)

Ramkumar, S.; Kirubakaran, V.

2016-01-01

Highlights: • The C.V of biodiesel is less than diesel of about 10% on volume and 15% on mass basis. • Most forms of biodiesel and its blends with diesel have higher viscosity than diesel. • Biodiesel’s cost and by-product reduce its feasibility as a substitute fuel. • TGA & DCS of Pungamia Oil shows that Thermal cracking is an alternate to Biodiesel. - Abstract: The awareness about using eco friendly fuels like biodiesel is increasing every day. The Increase in global warming and energy crises due to fossil fuel has accelerated the search of bio fuels. Biodiesel is a promising fuel; it is available in a wide range in every part of the world. Most of the studies reveal that the performance of biodiesel is better than that of diesel. Except NOx, the major emissions are high in the case of fossil fuels. This paper reviews the performance and emission characteristics of biodiesel in C.I engines. The paper also reviews the influence of engine modifications, various additives, and various proportions of blends of biodiesel with diesel. The physical and thermal characteristics of biodiesel have a great influence in the performance and emission, and they are tabulated in this paper. This paper also attempts feasibility of admitting vegetable oil in IC engine through Thermal Cracking. Preliminary investigation shows encouraging results and reported in this paper.

Thermal dimensioning of spent fuel repository

International Nuclear Information System (INIS)

Ikonen, K.

2009-09-01

This report contains the temperature dimensioning of the KBS-3V type nuclear fuel repository in Olkiluoto for the BWR, VVER and EPR fuel canisters, which are disposed at vertical position in the horizontal tunnels in a rectangular geometry according to the preliminary Posiva plan. This report concerns only the temperature dimensioning of the repository and does not take into account the possible restrictions caused by the stresses induced in the rock. The maximum temperature on the canister-bentonite interface is limited to the design temperature of +100 deg C. However, due to uncertainties in thermal analysis parameters (like scattering in rock conductivity or in predicted decay power) the allowable calculated maximum canister temperature is set to 90 deg C causing a safety margin of 10 deg C. The allowable temperature is controlled by adjusting the space between adjacent canisters, adjacent tunnels and the pre-cooling time affecting on power of the canisters. The temperature of canister surfaces can be determined by superposing analytic line heat source models much more efficiently than by numerical analysis, if the analytic model is first calibrated by numerical analysis (by control volume method). This was done by comparing the surface temperatures of a single canister calculated numerically and analytically. For the Olkiluoto repository of one panel having 900 canisters of BWR, VVER and EPR spent fuel was analyzed. The analyses were performed with an initial canister power of 1 700 W, 1 370 W and 1 830 W, respectively. These decay heats are obtained when the pre-cooling times of the fuels are 32.9, 29.6 and 50.3 years (the burn-up values 40, 40 and 50 MWd/kgU, respectively). The analyses gave as a result the canister spacing (6.0-10.8 m), when the tunnel spacing was 25 m, 30 m or 40 m. On the edge areas of the panel with constant canister spacing the temperatures of the canisters are lower than in the middle area of the repository. Thus it is possible to pack

Modeling the influence of interaction layer formation on thermal conductivity of U–Mo dispersion fuel

International Nuclear Information System (INIS)

Burkes, Douglas E.; Casella, Andrew M.; Huber, Tanja K.

2015-01-01

Highlights: • Hsu equation provides best thermal conductivity estimate of U–Mo dispersion fuel. • Simple model considering interaction layer formation was coupled with Hsu equation. • Interaction layer thermal conductivity is not the most important attribute. • Effective thermal conductivity is mostly influenced by interaction layer formation. • Fuel particle distribution also influences the effective thermal conductivity. - Abstract: The Global Threat Reduction Initiative Program continues to develop existing and new test reactor fuels to achieve the maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Currently, the program is focused on assisting with the development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix, porosity that forms during fabrication of the fuel plates or rods, and upon the concentration of the dispersed phase within the matrix. This paper develops and validates a simple model to study the influence of interaction layer formation, dispersed particle size, and volume fraction of dispersed phase in the matrix on the effective conductivity of the composite. The model shows excellent agreement with results previously presented in the literature. In particular, the thermal conductivity of the interaction layer does not appear to be as important in determining the effective conductivity of the composite, while formation of the interaction layer and subsequent consumption of the matrix reveals a rather significant effect. The effective thermal conductivity of the composite can be influenced by the dispersed particle distribution by minimizing interaction

A cermet fuel reactor for nuclear thermal propulsion

International Nuclear Information System (INIS)

Kruger, G.

1991-01-01

Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk

Thermal analysis of transportation packaging for nuclear spent fuel

International Nuclear Information System (INIS)

Akamatsu, Hiroshi; Taniuchi, Hiroaki

1989-01-01

Safety analysis of transportation packaging for nuclear spent fuel comprises structural, thermal, containment, shielding and criticality factors, and the safety of a packaging is verified by these analyses. In thermal analysis, the temperature of each part of the packaging is calculated under normal and accident test conditions. As an example of thermal analysis, the temperature distribution of a packaging being subjected to a normal test was calculated by the TRUMP code and compared with measured data. (author)

The promise and challenges of cermet fueled nuclear thermal propulsion reactors

International Nuclear Information System (INIS)

Brengle, R.G.; Harty, R.B.; Bhattacharyya, S.K.

1993-06-01

The use of cermet fuels in nuclear thermal propulsion systems was examined and the characteristics of systems using these fuel forms is discussed in terms of current mission and safety requirements. For use at high temperatures cermet fueled reactors utilize ceramic fuels with refractory metals as the matrix material. Cermet fueled reactors tend to be heavy when compared to concepts that utilize graphite as the fuel matrix because of the high density of the refractory metal matrix which makes up 20-40 percent of the total volume. On the positive side the metal matrix is strong and more resistant to loads from either the launch or flow induced vibration. The compatibility of the tungsten cermet with hydrogen is excellent and lifetimes of several hours is certainly achievable. Probably the biggest drawback to cermet nuclear thermal propulsion concepts is that the amount of actual data to support the theoretical conclusions is small. In fact there is no data under representative conditions of temperature, propellant and flux for the required fuel burnup. Although cermet systems appear to be attractive, the lack of fuel data at representative conditions does not allow reliable comparisons of cermet systems to systems where fuel data is available. 10 refs

Thermal test and analysis of a spent fuel storage cask

International Nuclear Information System (INIS)

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

1993-01-01

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

Nuclear-Thermal Analysis of Fully Ceramic Microencapsulated Fuel via Two-Temperature Homogenized Model

International Nuclear Information System (INIS)

Lee, Yoon Hee; Cho, Nam Zin

2013-01-01

The FCM fuel is based on a proven safety philosophy that has been utilized operationally in very high temperature reactors (VHTRs). However, the FCM fuel consists of TRISO particles randomly dispersed in SiC matrix. The high heterogeneity in composition leads to difficulty in explicit thermal calculation of such a fuel. Therefore, an appropriate homogenization model becomes essential. In this paper, we apply the two-temperature homogenized model to thermal analysis of an FCM fuel. The model was recently proposed in order to provide more realistic temperature profiles in the fuel element in VHTRs. We applied the two-temperature homogenized model to FCM fuel. The two-temperature homogenized model was obtained by particle transport Monte Carlo calculation applied to the pellet region consisting of many coated particles uniformly dispersed in SiC matrix. Since this model gives realistic temperature profiles in the pellet (providing fuel-kernel temperature and SiC matrix temperature distinctly), it can be used for more accurate neutronics evaluation such as Doppler temperature feedback. The transient thermal calculation may be performed also more realistically with temperature-dependent homogenized parameters in various scenarios

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.

Sensitivity of LWR fuel cycle costs to uncertainties in detailed thermal cross sections

International Nuclear Information System (INIS)

Ryskamp, J.M.; Becker, M.; Harris, D.R.

1979-01-01

Cross sections averaged over the thermal energy (< 1 or 2 eV) group have been shown to have an important economic role for light-water reactors. Cost implications of thermal cross section uncertainties at the few-group level were reported earlier. When it has been determined that costs are sensitive to a specific thermal-group cross section, it becomes desirable to determine how specific energy-dependent cross sections influence fuel cycle costs. Multigroup cross-section sensitivity coefficients vary with fuel exposure. By changing the shape of a cross section displayed on a view-tube through an interactive graphics system, one can compute the change in few-group cross section using the exposure dependent sensitivity coefficients. With the changed exposure dependent few-group cross section, a new fuel cycle cost is computed by a sequence of batch depletion, core analysis, and fuel batch cost code modules. Fuel cycle costs are generally most sensitive to cross section uncertainties near the peak of the hardened Maxwellian flux

Thermal Design for Extra-Terrestrial Regenerative Fuel Cell System

Science.gov (United States)

Gilligan, R.; Guzik, M.; Jakupca, I.; Bennett, W.; Smith, P.; Fincannon, J.

2017-01-01

The Advanced Exploration Systems (AES) Advanced Modular Power Systems (AMPS) Project is investigating different power systems for various lunar and Martian mission concepts. The AMPS Fuel Cell (FC) team has created two system-level models to evaluate the performance of regenerative fuel cell (RFC) systems employing different fuel cell chemistries. Proton Exchange Membrane fuel cells PEMFCs contain a polymer electrolyte membrane that separates the hydrogen and oxygen cavities and conducts hydrogen cations (protons) across the cell. Solid Oxide fuel cells (SOFCs) operate at high temperatures, using a zirconia-based solid ceramic electrolyte to conduct oxygen anions across the cell. The purpose of the modeling effort is to down select one fuel cell chemistry for a more detailed design effort. Figures of merit include the system mass, volume, round trip efficiency, and electrolyzer charge power required. PEMFCs operate at around 60 C versus SOFCs which operate at temperatures greater than 700 C. Due to the drastically different operating temperatures of the two chemistries the thermal control systems (TCS) differ. The PEM TCS is less complex and is characterized by a single pump cooling loop that uses deionized water coolant and rejects heat generated by the system to the environment via a radiator. The solid oxide TCS has its own unique challenges including the requirement to reject high quality heat and to condense the steam produced in the reaction. This paper discusses the modeling of thermal control systems for an extraterrestrial RFC that utilizes either a PEM or solid oxide fuel cell.

Prediction of deformation and hygro-thermal stresses distribution in PEM fuel cell vehicle using three-dimensional CFD model

Energy Technology Data Exchange (ETDEWEB)

Sadiq Al-Baghdadi, Maher A.R. [Fuel Cell Research Center, International Energy & Environment Foundation, Al-Najaf, P.O.Box 39 (Iraq)

2012-07-01

Durability is one of the most critical remaining issues impeding successful commercialization of broad PEM fuel cell transportation energy applications. Automotive fuel cells are likely to operate with neat hydrogen under load-following or load-levelled modes and be expected to withstand variations in environmental conditions, particularly in the context of temperature and atmospheric composition. In addition, they are also required to survive over the course of their expected operational lifetimes i.e., around 5,500 hrs, while undergoing as many as 30,000 startup/shutdown cycles. The damage mechanisms in a PEM fuel cell are accelerated by mechanical stresses arising during fuel cell assembly (bolt assembling), and the stresses arise during fuel cell running, because it consists of the materials with different thermal expansion and swelling coefficients. Therefore, in order to acquire a complete understanding of the damage mechanisms in the membrane, mechanical response under steady-state hygro-thermal stresses should be studied under real cell operating conditions and in real cell geometry (three-dimensional). In this work, full three-dimensional, non-isothermal computational fluid dynamics model of a PEM fuel cell has been developed to simulate the stresses inside the PEM fuel cell, which are occurring during fuel cell assembly (bolt assembling), and the stresses arise during fuel cell running due to the changes of temperature and relative humidity. A unique feature of the present model is to incorporate the effect of hygro and thermal stresses into actual three-dimensional fuel cell model. In addition, the temperature and humidity dependent material properties are utilize in the simulation for the membrane. The model is shown to be able to understand the many interacting, complex electrochemical, transport phenomena, and stresses distribution that have limited experimental data. This model is used to study and analyse the effect of operating parameters on the

High-fidelity simulations for clean and efficient combustion of alternative fuels

Energy Technology Data Exchange (ETDEWEB)

Oefelein, J C; Chen, J H [Reacting Flow Research Department, Sandia National Laboratories, Livermore, CA 94550 (United States); Sankaran, R, E-mail: oefelei@sandia.go [National Center for Computational Sciences, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2009-07-01

There is an urgent and growing demand for high-fidelity simulations that capture complex turbulence-chemistry interactions in propulsion and power systems, and in particular, that capture and discriminate the effects of fuel variability. This project addresses this demand using the Large Eddy Simulation (LES) technique (led by Oefelein) and the Direct Numerical Simulation (DNS) technique (led by Chen). In particular, we are conducting research under the INCITE program that is tightly coupled with funded projects established under the DOE Basic Energy Sciences and Energy Efficiency and Renewable Energy programs that will provide the foundational science required to develop a predictive modeling capability for design of advanced engines for transportation. Application of LES provides the formal ability to treat the full range of multidimensional time and length scales that exist in turbulent reacting flows in a computationally feasible manner and thus provides a way to simulate reacting flow phenomena in complex internal-combustion engine geometries at device relevant conditions. Application of DNS provides a way to study fundamental issues related to small-scale combustion processes in canonical configurations to understand dynamics that occur over a range of reactive-diffusive scales. Here we describe the challenges and present representative examples of the types of simulations each respective tool has been used for as part of the INCITE program. We focus on recent experiences on the Oak Ridge National Laboratory (ORNL) National Center for Computational Sciences (NCCS) Cray-XT Platform (i.e., Jaguar).

A cermet fuel reactor for nuclear thermal propulsion

Science.gov (United States)

Kruger, Gordon

1991-01-01

Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

Thermal hydraulic design of a hydride-fueled inverted PWR core

International Nuclear Information System (INIS)

Malen, J.A.; Todreas, N.E.; Hejzlar, P.; Ferroni, P.; Bergles, A.

2009-01-01

An inverted PWR core design utilizing U(45%, w/o)ZrH 1.6 fuel (here referred to as U-ZrH 1.6 ) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH 1.6 . The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW t , which is 135% of the optimally powered standard design (5080 MW t -determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.

Experimental Studies for CPF and SCR Model, Control System, and OBD Development for Engines Using Diesel and Biodiesel Fuels

Energy Technology Data Exchange (ETDEWEB)

Johnson, John; Naber, Jeffrey; Parker, Gordon; Yang, Song-Lin; Stevens, Andrews; Pihl, Josh

2013-04-30

The research carried out on this project developed experimentally validated Diesel Oxidation Catalyst (DOC), Diesel Particulate Filter (DPF), and Selective Catalytic Reduction (SCR) high‐fidelity models that served as the basis for the reduced order models used for internal state estimation. The high‐fidelity and reduced order/estimator codes were evaluated by the industrial partners with feedback to MTU that improved the codes. Ammonia, particulate matter (PM) mass retained, PM concentration, and NOX sensors were evaluated and used in conjunction with the estimator codes. The data collected from PM experiments were used to develop the PM kinetics using the high‐fidelity DPF code for both NO2 assisted oxidation and thermal oxidation for Ultra Low Sulfur Fuel (ULSF), and B10 and B20 biodiesel fuels. Nine SAE papers were presented and this technology transfer process should provide the basis for industry to improve the OBD and control of urea injection and fuel injection for active regeneration of the PM in the DPF using the computational techniques developed. This knowledge will provide industry the ability to reduce the emissions and fuel consumption from vehicles in the field. Four MS and three PhD Mechanical Engineering students were supported on this project and their thesis research provided them with expertise in experimental, modeling, and controls in aftertreatment systems.

Comprehensive thermal-hydraulic and thermal-mechanical analysis of core and fuel rods for the safety validation of real refueling at the Kozloduy WWER-440

Energy Technology Data Exchange (ETDEWEB)

Stefanova, S; Panajotov, D; Ilieva, B; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

Safety analysis aimed at determination of thermal-hydraulic and thermal-mechanical margins of core and fuel rods has been carried out using computer codes COBSOFM and PIN-micro. Thermal-hydraulic calculations for the part of the core with maximum heat flux during steady-state regime show that the coolant, cladding and fuel temperatures are within the design limits. A severe accident with reactor blackout has been simulated. It is found that at 95% probability level there is no boiling crisis anywhere in the core. The thermal-mechanical parameters of working assembly fuel rod with maximum load have been calculated. The assembly linear power reached a maximum of 25 kW/m during the second fuel cycle, the fuel temperature remaining well below 1000{sup o} C. As the fuel assembly with typical power history has enough safety margins, it was proposed to use it for one more cycle. 4 refs., 12 figs.

Thermal analysis of the IDENT 1578 fuel pin shipping container

International Nuclear Information System (INIS)

Ingham, J.G.

1980-01-01

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

Conceptual assessment and thermal hydraulic analysis of MVDS system for the dry storage of reduced metal fuel

International Nuclear Information System (INIS)

Lee, J. C.; Bang, K. S.; Shin, H. S.; Joo, J. S.; Su, K. S.; Kim, H. D.

2003-01-01

Conceptual assessment and thermal hydraulic analysis of MVDS storage system have been carried out for application of reduced metal fuel. The storage concept was established considering the optimum weight, storage volume and thermal efficiency. The capacity of MVDS system for loading the reduced metal fuel has four times as compared with existing PWR fuel storage system. In the results of thermal analysis, the maximum temperature of metal fuel was estimated to be 110 .deg. C which is lower than the allowable value under normal operation condition. Therefore, it is shown that the MVDS system can feasibly accomodate the reduced metal fuel in aspect of thermal safety

Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

Science.gov (United States)

Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

2014-12-01

Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

Status and prospects of thermal breeders and their effect on fuel utilization

Energy Technology Data Exchange (ETDEWEB)

1979-01-01

The report evaluates the extent to which thermal breeders and near-breeders might complement fast breeders or serve as an alternative in solving the long-term nuclear fuel supply problem. It considers in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability and fuel cycle versatility in order to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues.

Status and prospects of thermal breeders and their effect on fuel utilization

International Nuclear Information System (INIS)

1979-01-01

The report evaluates the extent to which thermal breeders and near-breeders might complement fast breeders or serve as an alternative in solving the long-term nuclear fuel supply problem. It considers in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability and fuel cycle versatility in order to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

Evaluation of Candidate In-Pile Thermal Conductivity Techniques

International Nuclear Information System (INIS)

Fox, B.; Ban, H.; Daw, J.; Condie, K.; Knudson, D.; Rempe, J.

2009-01-01

Thermophysical properties of materials must be known for proper design, test, and application of new fuels and structural properties in nuclear reactors. In the case of nuclear fuels during irradiation, the physical structure and chemical composition change as a function of time and position within the rod. Typically, thermal conductivity changes, as well as other thermophysical properties being evaluated during irradiation in a materials and test reactor, are measured out-of-pile in 'hot-cells'. Repeatedly removing samples from a test reactor to make out-of-pile measurements is expensive, has the potential to disturb phenomena of interest, and only provide understanding of the sample's end state at the time each measurement is made. There are also limited thermophysical property data for advanced fuels. Such data are needed for the development of next generation reactors and advanced fuels for existing nuclear plants. Having the capacity to effectively and quickly characterize fuels and material properties during irradiation has the potential to improve the fidelity of nuclear fuel data and reduce irradiation testing costs

Thermal behaviour of high burnup PWR fuel under different fill gas conditions

International Nuclear Information System (INIS)

Tverberg, T.

2001-01-01

During its more than 40 years of existence, a large number of experiments have been carried out at the Halden Reactor Project focusing on different aspects related to nuclear reactor fuel. During recent years, the fuels testing program has mainly been focusing on aspects related to high burnup, in particular in terms of fuel thermal performance and fission gas release, and often involving reinstrumentation of commercially irradiated fuel. The paper describes such an experiment where a PWR rod, previously irradiated in a commercial reactor to a burnup of ∼50 MWd/kgUO 2 , was reinstrumented with a fuel central oxide thermocouple and a cladding extensometer together with a high pressure gas flow line, allowing for different fill gas compositions and pressures to be applied. The paper focuses on the thermal behaviour of such LWR rods with emphasis on how different fill gas conditions influence the fuel temperatures and gap conductance. Rod growth rate was also monitored during the irradiation in the Halden reactor. (author)

Thermal Management Tools for Propulsion System Trade Studies and Analysis

Science.gov (United States)

McCarthy, Kevin; Hodge, Ernie

2011-01-01

Energy-related subsystems in modern aircraft are more tightly coupled with less design margin. These subsystems include thermal management subsystems, vehicle electric power generation and distribution, aircraft engines, and flight control. Tighter coupling, lower design margins, and higher system complexity all make preliminary trade studies difficult. A suite of thermal management analysis tools has been developed to facilitate trade studies during preliminary design of air-vehicle propulsion systems. Simulink blocksets (from MathWorks) for developing quasi-steady-state and transient system models of aircraft thermal management systems and related energy systems have been developed. These blocksets extend the Simulink modeling environment in the thermal sciences and aircraft systems disciplines. The blocksets include blocks for modeling aircraft system heat loads, heat exchangers, pumps, reservoirs, fuel tanks, and other components at varying levels of model fidelity. The blocksets have been applied in a first-principles, physics-based modeling and simulation architecture for rapid prototyping of aircraft thermal management and related systems. They have been applied in representative modern aircraft thermal management system studies. The modeling and simulation architecture has also been used to conduct trade studies in a vehicle level model that incorporates coupling effects among the aircraft mission, engine cycle, fuel, and multi-phase heat-transfer materials.

Thermal stress analysis of sulfur deactivated solid oxide fuel cells

Science.gov (United States)

Zeng, Shumao; Parbey, Joseph; Yu, Guangsen; Xu, Min; Li, Tingshuai; Andersson, Martin

2018-03-01

Hydrogen sulfide in fuels can deactivate catalyst for solid oxide fuel cells, which has become one of the most critical challenges to stability. The reactions between sulfur and catalyst will cause phase changes, leading to increase in cell polarization and mechanical mismatch. A three-dimensional computational fluid dynamics (CFD) approach based on the finite element method (FEM) is thus used to investigate the polarization, temperature and thermal stress in a sulfur deactivated SOFC by coupling equations for gas-phase species, heat, momentum, ion and electron transport. The results indicate that sulfur in fuels can strongly affect the cell polarization and thermal stresses, which shows a sharp decrease in the vicinity of electrolyte when 10% nickel in the functional layer is poisoned, but they remain almost unchanged even when the poisoned Ni content was increased to 90%. This investigation is helpful to deeply understand the sulfur poisoning effects and also benefit the material design and optimization of electrode structure to enhance cell performance and lifetimes in various hydrocarbon fuels containing impurities.

Modeling the Thermal Rocket Fuel Preparation Processes in the Launch Complex Fueling System

Directory of Open Access Journals (Sweden)

A. V. Zolin

2015-01-01

hydrocarbon fuel returning to the storage tank.Mathematical models of cooling and heating processes are built on the assumption that the heat exchange process of storage and environment is quasistationary.The paper presents relationships for determining the relative masses of nitrogen and time to perform the operation of cooling fuel from the initial to the desired final temperature as well as relationships to define the time of heating operation for a given capacity of the heat exchanger-heater and the pump station fueling system.The results of calculations of the relative liquid nitrogen costs during cooling of hydrocarbon gases depending on the mass flow rate of nitrogen in the cooling fuel system are shown in comparison with experimental data and numerical calculations. The maximum error of analytical calculation results and experimental values of the relative cost of liquid nitrogen does not exceed 4.5% and the error in determining the time required for operations of temperature preparation does not exceed 5%.Analytical relationships and results of calculations obtained on their basis are adequate and in compliance with experimental results, in accuracy are on a par with results of numerical calculations and, as compared to numerical solution, greatly simplify a procedure of implemented design calculations of fuel temperature preparation processes. Using these relationships allows to analyze the effectiveness of the operations of heating and cooling hydrocarbon fuel depending on the design parameters of the storage capacity, its thermal insulation, mass of fuel, thermal power of the heating devices, flow of nitrogen, as well as to determine the required mass of liquid nitrogen and the operation parameters of cooling (heating fuel for filling systems of launch complexes for different values of the environmental parameters, the initial and desired final temperaturesof the fuel.

Probabilistic analysis of strength and thermal-physic WWER fuel rod characteristics using START-3 code

International Nuclear Information System (INIS)

Medvedev, A.; Bogatyr, S.; Khramtsov; Sokolov, F.

2001-01-01

During the last years probabilistic methods for evaluation of the influence of the fuel geometry and technology parameters on fuel operational reliability are widely used. In the present work the START-3 procedure is used to calculate the thermal physics and strength characteristics of WWER fuel rods behavior. The procedure is based on the Monte-Carlo method with the application of Sobol quasi-random sequences. This technique allows to treat the fuel rod technological and operating parameters as well as its strength and thermal physics characteristics as random variables. The work deals with a series of WWER-1000 fuel rod statistical tests and verification based on the PIE results. Also preliminary calculations are implemented with the aim to determine the design schema parameters. This should ensure the accuracy of the assessment of the parameters of WWER fuel rod characteristics distribution. The probability characteristics of fuel rod strength and thermal physics are assessed via the statistical analysis of the results of probability calculations

Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

International Nuclear Information System (INIS)

Waata, C.L.

2006-07-01

The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Bodey, Isaac T [ORNL

2014-05-01

Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

Effect of 17 x 17 fuel assembly geometry on interchannel thermal mixing

International Nuclear Information System (INIS)

Motley, F.E.; Wenzell, A.H.; Cadek, F.F.

1975-01-01

A test to determine the value of the thermal diffusion coefficient (TDC) in the 17 x 17 fuel assembly geometry was conducted. The test section was a 5 x 5 rod bundle with a radial power difference of 4.5 to 1. The rod OD and pitch are identical to the 17 x 17 fuel assembly, as is the mixing vane grid design. The value of thermal diffusion coefficient (TDC) was determined by matching the experimental exit enthalpy distribution to that predicted by the THINC computer code. The mean value of TDC for the 17 x 17 fuel assembly geometry is TDC = .059. 6 references

Thermal analysis of a one-element PWR spent fuel shipping cask

International Nuclear Information System (INIS)

Fields, S.R.

1979-06-01

The transient thermal behavior of a typical one-element PWR spent fuel shipping cask, following a hypothetical accident and fire, has been simulated. The objectives of the study were to determine the transient behavior of the cask and its spent fuel primary coolant through the pressure relief system and possible fuel pin clad failure due to overheating following loss of coolant. 15 figures, 7 tables

Thermal analyses of spent nuclear fuel repository

International Nuclear Information System (INIS)

Ikonen, K.

2003-06-01

This report contains the temperature dimensioning of the KBS-3V type 1- or 2-panel repository based on the rock properties measured from the Olkiluoto investigations. The report describes first the development of a calculation methodology for the thermal analysis of a repository for nuclear fuel. The disposed canisters produce residual heat due to decay (or disintegration) of radioactive products. The decay heat is conducted to surrounding rock mass. The methods were applied to determine the effect of different parameters on the highest canister temperature and to support the planning, dimensioning and operation of the repository. The thermal diffusivity of the rock is low and the heat released from the canisters is spread into the surrounding rock volume quite slowly causing thermal gradient in the rock close to canisters and the canister temperature is increased remarkably. The maximum temperature on the canister surface is limited to the design temperature of +100 deg C. However, due to uncertainties in thermal analysis parameters (like scattering in rock conductivity) the allowable calculated maximum canister temperature is set to 90 deg C causing a safety margin of 10 deg C. The allowable temperature is controlled by the spacing between adjacent canisters, adjacent tunnels and the distance between separate panels of the repository and the pre-cooling time affecting power of the canisters. Because of the fact that the disposal operation takes several decades, the moment of disposal of an individual canister in addition to the location has an influence on the maximum temperature in the canister. Also, a second disposal panel in the repository has a thermal interaction with the other panel. This interaction is expressed after a few decades at the strongest. It became apparent that the temperature of canister surfaces can be determined by analytic line heat source model much more efficiently than by numerical analysis, if the analytic model is first verified and

TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

International Nuclear Information System (INIS)

Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

1983-01-01

TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

CFD analysis of aircraft fuel tanks thermal behaviour

Science.gov (United States)

Zilio, C.; Longo, G. A.; Pernigotto, G.; Chiacchio, F.; Borrelli, P.; D'Errico, E.

2017-11-01

This work is carried out within the FP7 European research project TOICA (Thermal Overall Integrated Conception of Aircraft, http://www.toica-fp7.eu/). One of the tasks foreseen for the TOICA project is the analysis of fuel tanks as possible heat sinks for future aircrafts. In particular, in the present paper, commercial regional aircraft is considered as case study and CFD analysis with the commercial code STAR-CCM+ is performed in order to identify the potential capability to use fuel stored in the tanks as a heat sink for waste heat dissipated by other systems. The complex physical phenomena that characterize the heat transfer inside liquid fuel, at the fuel-ullage interface and inside the ullage are outlined. Boundary conditions, including the effect of different ground and flight conditions, are implemented in the numerical simulation approach. The analysis is implemented for a portion of aluminium wing fuel tank, including the leading edge effects. Effect of liquid fuel transfer among different tank compartments and the air flow in the ullage is included. According to Fuel Tank Flammability Assessment Method (FTFAM) proposed by the Federal Aviation Administration, the results are exploited in terms of exponential time constants and fuel temperature difference to the ambient for the different cases investigated.

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

Fuel Thermo-physical Characterization Project: Evaluation of Models to Calculate Thermal Diffusivity of Layered Composites

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gardner, Levi D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Huber, Tanja K. [Technische Universität München, Munich (Germany); Breitkreutz, Harald [Technische Universität München, Munich (Germany)

2015-02-11

The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universität München (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

A comparison of thermal algorithms of fuel rod performance code systems

International Nuclear Information System (INIS)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

A comparison of thermal algorithms of fuel rod performance code systems

Energy Technology Data Exchange (ETDEWEB)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

Thermally-Constrained Fuel-Optimal ISS Maneuvers

Science.gov (United States)

Bhatt, Sagar; Svecz, Andrew; Alaniz, Abran; Jang, Jiann-Woei; Nguyen, Louis; Spanos, Pol

2015-01-01

Optimal Propellant Maneuvers (OPMs) are now being used to rotate the International Space Station (ISS) and have saved hundreds of kilograms of propellant over the last two years. The savings are achieved by commanding the ISS to follow a pre-planned attitude trajectory optimized to take advantage of environmental torques. The trajectory is obtained by solving an optimal control problem. Prior to use on orbit, OPM trajectories are screened to ensure a static sun vector (SSV) does not occur during the maneuver. The SSV is an indicator that the ISS hardware temperatures may exceed thermal limits, causing damage to the components. In this paper, thermally-constrained fuel-optimal trajectories are presented that avoid an SSV and can be used throughout the year while still reducing propellant consumption significantly.

Thermal stability and filterability of jet fuels containing PDR additives in small-scale tests and realistic rig simulations

Energy Technology Data Exchange (ETDEWEB)

Bauldreay, J.M.; Clark, R.H.; Heins, R.J. [Shell Research, Ltd., Chester (United Kingdom)

1995-05-01

Specification, small-scale and realistic fuel simulation tests have addressed concerns about the impact of pipeline drag reducer (PDR) flow modifying additives on jet fuel handling and performance. A typical PDR additive tended to block filters which were similar to those used in the specification Jet Fuel Thermal Oxidation Tester (JFTOT) and other thermal stability test apparatus. Blockages reduced flow rates and PDR concentrations downstream of the filters. Consequently two PDR additives (A&B) were tested in JFTOT apparatus without the usual in-line pre-filters as part of a Ministry of Defense (MoD) co-ordinated Round Robin exercise. Some fuel/PDR additive combinations caused decreases in JFTOT breakpoints. Effects were additive- (type, concentration and degree of shear) and fuel-dependent; most failures were caused by filter blockages and not by a failing lacquer rating. In further work at Thornton, the thermal stability characteristics of similar fuel/additive combinations have been examined in non-specification tests. In Flask Oxidation Tests, PDR additives caused no significant increase in the liquid phase oxidation rates of the fuels. Additives were tested in the Single Tube Heat Transfer Rig (STHTR) which duplicates many of the conditions of a heat exchanger element in an engine`s fuel supply system. B produced an average two-fold decrease in thermal stability in a Merox fuel; A had no significant effect. In hydrotreated fuel, B reduced the thermal stability up to five-fold. A had little effect below 205{degrees}C, while at higher temperatures there may have been a marginal improvement in thermal stability. Again, certain jet fuel/PDR combinations were seen to reduce thermal stability.

Advances in estimation technology of thermal conductivity of irradiated fuels (1). Application of a thermal microscope to measure the thermal conductivity of the second phases in irradiated pellets

International Nuclear Information System (INIS)

Uno, Masayoshi; Murakami, Yukihiro

2011-01-01

CeO 2 sample as a surrogate for fuel and BaCeO 3 and BaMoO 4 samples as surrogates for the second phases, which have a lower thermal conductivity than the fuel matrix, were made. The thermal conductivity of these samples was measured by a thermal microscope. In this method, the thermal conductivity of a small region (e.g. 20 μm x 20 μm) of the sample can be measured. The valid thermal conductivity values for all the samples were obtained and the conditions of sample surface preparation and the thermal microscope measurement were found out. The thermal conductivity of a CeO 2 composite pellet which had the BaCeO 3 or BaMoO 4 second phase layer was also estimated. (author)

Increase of thermal conductivity of uranium dioxide nuclear fuel pellets with beryllium oxide addition

International Nuclear Information System (INIS)

Camarano, D.M.; Mansur, F.A.; Santos, A.M.M. dos; Ferraz, W.B.

2016-01-01

The UO_2 fuel is one of the most used nuclear fuel in thermal reactors and has many advantages such as high melting point, chemical compatibility with cladding, etc. However, its thermal conductivity is relatively low, which leads to a premature degradation of the fuel pellets due to a high radial temperature gradient during reactor operation. An alternative to avoid this problem is to increase the thermal conductivity of the fuel pellets, by adding beryllium oxide (BeO). Pellets of UO_2 and UO_2-BeO were obtained from a homogenized mixture of powders of UO_2 and BeO, containing 2% and 3% by weight of BeO and sintering at 1750 °C for 3 h under H_2 atmosphere after uniaxial pressing at 400 MPa. The pellet densities were obtained by xylol penetration-immersion method and the thermal diffusivity, specific heat and thermal conductivity were determined according to ASTM E-1461 at room temperature (25 deg C) and 100 deg C. The thermal diffusivity measurements were carried out employing the laser flash method. The thermal conductivity obtained at 25 deg C showed an increase with the addition of 2% and 3% of BeO corresponding to 19% and 28%, respectively. As for the measurements carried out at 100 deg C, there was an increase in the thermal conductivity for the same BeO contents of 20% and 31%. These values as a percentage of increased conductivity were obtained in relation to the UO_2 pellets. (author)

Development of Innovative Accident Tolerant High Thermal Conductivity UO₂-Diamond Composite Fuel Pellets

Energy Technology Data Exchange (ETDEWEB)

Tulenko, James [Univ. of Florida, Gainesville, FL (United States); Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States)

2016-01-01

The University of Florida (UF) evaluated a composite fuel consisting of UO₂ powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO₂-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO₂ – SiC and UO₂ – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO₂ may greatly enhanced the thermal conductivity of the UO₂ pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

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

International Nuclear Information System (INIS)

Copsey, A.B.

1992-01-01

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

Thermal hydraulic feasibility assessment of the spent nuclear fuel project

International Nuclear Information System (INIS)

Heard, F.J.

1996-01-01

A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The goal was to develop a series of thermal-hydraulic models that could respond to all process and safety related issues that may arise pertaining to the SNFP, as well as provide a basis for validation of the results. Results show that there is a reasonable envelope for process conditions and requirements that are thermally and hydraulically acceptable

Thermal Analysis of Solid Fuels in an Inert Atmosphere

Science.gov (United States)

Kijo-Kleczkowska, Agnieszka; Szumera, Magdalena; Środa, Katarzyna

2017-12-01

The paper takes the analysis of thermal studies of different types of fuels. It allowed diversification of fuels depending on their composition and origin. Consideration of coal, biomass and waste (coal mule, sewage sludge) as fuel is nowadays an important aspect of energy in our country. It should be emphasized that Poland power engineering is based up to 95% on coal - the primary fuel. Mining industry, forced to deliver power engineering more and better fuel, must however, use a deeper cleaning of coal. This results in a continuous increase waste in the form of mule flotation. The best method of disposing these mule is combustion and co-combustion with other fuels. On the other hand, commonly increasing awareness state of the environment and the need to reduce CO2 emissions energy industry have committed to implement alternative solutions in order to gain power, through, i.a.: development technologies use of biomass, which is one of the most promising renewable energy sources in Poland. The paper presents the results of research TG-DTA fuels made in an inert atmosphere.

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

International Nuclear Information System (INIS)

Daw, J.E.; Knudson, D.L.; Villard, J.F.; Liothin, J.; Destouches, C.; Rempe, J.L.; Matheron, P.; Lambert, T.

2015-01-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

Energy Technology Data Exchange (ETDEWEB)

Daw, J.E.; Knudson, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415, (United States); Villard, J.F.; Liothin, J.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Rempe, J.L. [Rempe and Associates, LLC, Idaho Falls, ID, 83404 (United States); Matheron, P. [CEA, DEN, DEC, Uranium Fuels Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Lambert, T. [CEA, DEN, DEC, Innovative Fuel Design and Irradiation Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France)

2015-07-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

Thorium utilization: conversion ratio and fuel needs in thermal reactors

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1975-01-01

As a preparatory study for thorium utilization in thermal reactors a study has been made of the fuel comsumption in existing reactor types. A quantitative description is given of the influence of enrichment, burnup, amount of structural material, choise of coolant and control requirements on the convertion ratio. The enrichment is an important factor and a low fuel comsumption can be achieved by increasing the enrichment

Three-dimensional fuel pin model validation by prediction of hydrogen distribution in cladding and comparison with experiment

Energy Technology Data Exchange (ETDEWEB)

Aly, A. [North Carolina State Univ., Raleigh, NC (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States); Ivanov, Kostadin [Pennsylvania State Univ., University Park, PA (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Lacroix, E. [Pennsylvania State Univ., University Park, PA (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Walter, D. [Univ. of Michigan, Ann Arbor, MI (United States); Williamson, R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gamble, K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-10-29

To correctly describe and predict this hydrogen distribution there is a need for multi-physics coupling to provide accurate three-dimensional azimuthal, radial, and axial temperature distributions in the cladding. Coupled high-fidelity reactor-physics codes with a sub-channel code as well as with a computational fluid dynamics (CFD) tool have been used to calculate detailed temperature distributions. These high-fidelity coupled neutronics/thermal-hydraulics code systems are coupled further with the fuel-performance BISON code with a kernel (module) for hydrogen. Both hydrogen migration and precipitation/dissolution are included in the model. Results from this multi-physics analysis is validated utilizing calculations of hydrogen distribution using models informed by data from hydrogen experiments and PIE data.

Thermal conductivity of U–Mo/Al dispersion fuel. Effects of particle shape and size, stereography, and heat generation

International Nuclear Information System (INIS)

Cho, Tae Won; Sohn, Dong-Seong; Kim, Yeon Soo

2015-01-01

This paper describes the effects of particle sphericity, interfacial thermal resistance, stereography, and heat generation on the thermal conductivity of U–Mo/Al dispersion fuel. The ABAQUS finite element method (FEM) tool was used to calculate the effective thermal conductivity of U–Mo/Al dispersion fuel by implementing fuel particles. For U–Mo/Al, the particle sphericity effect was insignificant. However, if the effect of the interfacial thermal resistance between the fuel particles and Al matrix was considered, the thermal conductivity of U–Mo/Al was increased as the particle size increases. To examine the effect of stereography, we compared the two-dimensional modeling and three-dimensional modeling. The results showed that the two-dimensional modeling predicted lower than the three-dimensional modeling. We also examined the effect of the presence of heat sources in the fuel particles and found a decrease in thermal conductivity of U–Mo/Al from that of the typical homogeneous heat generation modeling. (author)

Advanced thermally stable jet fuels. Technical progress report, October 1993--December 1993

Energy Technology Data Exchange (ETDEWEB)

Schobert, H.H.; Eser, S.; Song, C.; Hatcher, P.G.; Walsh, P.M.; Coleman, M.M.

1994-01-01

The Penn State program in advancd thermally stable coal-based jet fuels has five broad objectives: (1) development of mechanisms of degradation and solids formation; (2) quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) elucidation of the role of additives in retarding them formation of vcarbonaceous solids; and, (5) assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal.

Thermal durability of modified Synroc material as reactor fuel matrix

International Nuclear Information System (INIS)

Kikuchi, Akira; Kanazawa, Hiroyuki; Togashi, Yoshihiro; Matumoto, Seiichiro; Nishino, Yasuharu; Ohwada, Isao; Nakata, Masahito; Amano, Hidetoshi; Mitamura, Hisayoshi

1994-08-01

A Synroc, a polyphase titanate ceramics composed of three mineral phases (perovskite, hollandite and zirconolite), has an excellent performance of immobilization of high level nuclear waste. A working group in the Department of Hot Laboratories paid special attention to this merit and started a development study on a LWR fuel named 'Waste Disposal Possible (WDP) Fuel', which has the two functions of a reactor fuel and a waste form. The present paper mainly describes thermal durability of a modified Synroc material, which is essentially important for applying the material to a fuel matrix. The two kinds of Synroc specimens, designated 'SM' as modified and 'SB' as a reference, were prepared by hot-pressing and annealed at 1200degC to 1500degC for 30 min in air. Unexpected and peculiar spherical voids were observed in the specimen SM at 1400degC and 1500degC, which caused the specimen swelling. The formation of the voids depends significantly on the existence of spherical precipitates seen in the as-fabricated specimen including latent micropores with high pressure. On the other hand, the heat treatment at 1500degC formed additional new phases, designated 'Phase A' for the specimen SB and 'Phase X' for SM. Phase A is a decomposition product of hollandite and Phase X a reaction product of Phase A and perovskite in the spherical voids. Furthermore, additional information and thermal properties examined are presented in Appendix 1 and Appendix 2, respectively. It was recognized that the modified Synroc specimen SM had excellent thermal properties. (author)

Combustion, Performance, and Emission Evaluation of a Diesel Engine with Biodiesel Like Fuel Blends Derived From a Mixture of Pakistani Waste Canola and Waste Transformer Oils

Directory of Open Access Journals (Sweden)

Muhammad Qasim

2017-07-01

Full Text Available The aim of this work was to study the combustion, performance, and emission characteristics of a 5.5 kW four-stroke single-cylinder water-cooled direct-injection diesel engine operated with blends of biodiesel-like fuel (BLF15, BLF20 & BLF25 obtained from a 50:50 mixture of transesterified waste transformer oil (TWTO and waste canola oil methyl esters (WCOME with petroleum diesel. The mixture of the waste oils was named as biodiesel-like fuel (BLF.The engine fuelled with BLF blends was evaluated in terms of combustion, performance, and emission characteristics. FTIR analysis was carried out to know the functional groups in the BLF fuel. The experimental results revealed the shorter ignition delay and marginally higher brake specific fuel consumption (BSFC, brake thermal efficiency (BTE and exhaust gas temperature (EGT values for BLF blends as compared to diesel. The hydrocarbon (HC and carbon monoxide (CO emissions were decreased by 10.92–31.17% and 3.80–6.32%, respectively, as compared to those of diesel fuel. Smoke opacity was significantly reduced. FTIR analysis has confirmed the presence of saturated alkanes and halide groups in BLF fuel. In comparison to BLF20 and BLF25, the blend BLF15 has shown higher brake thermal efficiency and lower fuel consumption values. The HC, CO, and smoke emissions of BLF15 were found lower than those of petroleum diesel. The fuel blend BLF15 is suggested to be used as an alternative fuel for diesel engines without any engine modification.

Study on the thermal-hydraulic stability of high burn up STEP III fuel in Japan

International Nuclear Information System (INIS)

Ishikawa, M.; Kitamura, H.; Toba, A.; Omoto, A.

2004-01-01

Japanese BWR utilities have performed a joint study of the Thermal Hydraulic Stability of High Burn up STEP III Fuel. In this study, the parametric dependency of thermal hydraulic stability threshold was obtained. It was confirmed through experiments that the STEP III Fuel has sufficient stability characteristics. (author)

Three-dimensional thermal analysis of a baseline spent fuel repository

International Nuclear Information System (INIS)

Altenbach, T.J.; Lowry, W.E.

1980-01-01

A three-dimensional thermal analysis has been performed using finite difference techniques to determine the near-field response of a baseline spent fuel repository in a deep geologic salt medium. A baseline design incorporates previous thermal modeling experience and OWI recommendations for areal thermal loading in specifying the waste form properties, package details, and emplacement configuration. The base case in this thermal analysis considers one 10-year old PWR spent fuel assembly emplaced to yield a 36 kw/acre (8.9 w/m 2 ) loading. A unit cell model in an infinite array is used to simplify the problem and provide upper-bound temperatures. Boundary conditions are imposed which allow simulations to 1000 years. Variations studied include a comparison of ventilated and unventilated storage room conditions, emplacement packages with and without air gaps surrounding the canister, and room cool-down scenarios with ventilation following an unventilated state for retrieval purposes. At this low power level ventilating the emplacement room has an immediate cooling influence on the canister and effectively maintains the emplacement room floor near the temperature of the ventilating air. The annular gap separating the canister and sleeve causes the peak temperature of the canister surface to rise by 10 0 F (5.6 0 C) over that from a no gap case assuming perfect thermal contact. It was also shown that the time required for the emplacement room to cool down to 100 0 F (38 0 C) from an unventilated state ranged from 2 weeks to 6 months; when ventilation initiated after times of 5 years to 50 years, respectively. As the work was performed for the Nuclear Regulatory Commission, these results provide a significant addition to the regulatory data base for spent fuel performance in a geologic repository

High-Fidelity Multidisciplinary Design Optimization of Aircraft Configurations

Science.gov (United States)

Martins, Joaquim R. R. A.; Kenway, Gaetan K. W.; Burdette, David; Jonsson, Eirikur; Kennedy, Graeme J.

2017-01-01

To evaluate new airframe technologies we need design tools based on high-fidelity models that consider multidisciplinary interactions early in the design process. The overarching goal of this NRA is to develop tools that enable high-fidelity multidisciplinary design optimization of aircraft configurations, and to apply these tools to the design of high aspect ratio flexible wings. We develop a geometry engine that is capable of quickly generating conventional and unconventional aircraft configurations including the internal structure. This geometry engine features adjoint derivative computation for efficient gradient-based optimization. We also added overset capability to a computational fluid dynamics solver, complete with an adjoint implementation and semiautomatic mesh generation. We also developed an approach to constraining buffet and started the development of an approach for constraining utter. On the applications side, we developed a new common high-fidelity model for aeroelastic studies of high aspect ratio wings. We performed optimal design trade-o s between fuel burn and aircraft weight for metal, conventional composite, and carbon nanotube composite wings. We also assessed a continuous morphing trailing edge technology applied to high aspect ratio wings. This research resulted in the publication of 26 manuscripts so far, and the developed methodologies were used in two other NRAs. 1

Cna 1 spent fuel element interim dry storage system thermal analysis

International Nuclear Information System (INIS)

Hilal, R. E; Garcia, J. C; Delmastro, D. F

2006-01-01

At the moment, the Atucha I Nuclear Power Plant (Cnea-I) located in the city of Lima, has enough room to store its spent fuel (Sf) in their two pools spent fuel until about 2015.In case of life extension a spend fuel element interim dry storage system is needed.Nucleolectrica Argentina S.A. (N A-S A) and the Comision Nacional de Energia Atomica (Cnea), have proposed different interim dry storage systems.These systems have to be evaluated in order to choose one of them.The present work's objective is the thermal analysis of one dry storage alternative for the Sf element of Cna 1.In this work a simple model was developed and used to perform the thermal calculations corresponding to the system proposed by Cnea.This system considers the store of sealed containers with 37 spent fuels in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.Fulfill the maximum cladding temperature requirement ( [es

Steady- and transient-state analyses of fully ceramic microencapsulated fuel loaded reactor core via two-temperature homogenized thermal-conductivity model

International Nuclear Information System (INIS)

Lee, Yoonhee; Cho, Nam Zin

2015-01-01

Highlights: • Fully ceramic microencapsulated fuel-loaded core is analyzed via a two-temperature homogenized thermal-conductivity model. • The model is compared to harmonic- and volumetric-average thermal conductivity models. • The three thermal analysis models show ∼100 pcm differences in the k eff eigenvalue. • The three thermal analysis models show more than 70 K differences in the maximum temperature. • There occur more than 3 times differences in the maximum power for a control rod ejection accident. - Abstract: Fully ceramic microencapsulated (FCM) fuel, a type of accident-tolerant fuel (ATF), consists of TRISO particles randomly dispersed in a SiC matrix. In this study, for a thermal analysis of the FCM fuel with such a high heterogeneity, a two-temperature homogenized thermal-conductivity model was applied by the authors. This model provides separate temperatures for the fuel-kernels and the SiC matrix. It also provides more realistic temperature profiles than those of harmonic- and volumetric-average thermal conductivity models, which are used for thermal analysis of a fuel element in VHTRs having a composition similar to the FCM fuel, because such models are unable to provide the fuel-kernel and graphite matrix temperatures separately. In this study, coupled with a neutron diffusion model, a FCM fuel-loaded reactor core is analyzed via a two-temperature homogenized thermal-conductivity model at steady- and transient-states. The results are compared to those from harmonic- and volumetric-average thermal conductivity models, i.e., we compare k eff eigenvalues, power distributions, and temperature profiles in the hottest single-channel at steady-state. At transient-state, we compare total powers, reactivity, and maximum temperatures in the hottest single-channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized thermal

The effect of copper, MDA, and accelerated aging on jet fuel thermal stability as measured by the gravimetric JFTOT

Energy Technology Data Exchange (ETDEWEB)

Pande, S.G. [Geo-Centers, Inc., Ft. Washington, MD (United States); Hardy, D.R. [Navy Technology Center for Safety and Survivability, Washington, DC (United States)

1995-05-01

Thermally unstable jet fuels pose operational problems. In order to adequately identify such fuels, factors that realistically impact on thermal stability were examined. Evaluation was based on a quantitative method of measuring thermal stability, viz., NRL`s recently developed gravimetric JFTOT. This method gives a quantitative measurement of both the strip deposit and filterables formed. The pertinent factors examined, included the individual and interactive effects of: soluble copper, MDA (metal deactivator), and aging. The latter was accelerated to simulate field conditions of approximately six months aging at ambient temperature and pressure. The results indicate that the individual and interactive effects of copper, MDA, and accelerated aging appear to be fuel dependent. Based on the results, the three test fuels examined (one JP-8 and two JP-5s) were categorized as exhibiting very good, typical, and poor thermal stabilities, respectively. For both the very good and poor thermal stability fuels, the effect of copper in conjunction with accelerated aging did not significantly increase the total thermal deposits of the neat fuels. In contrast, for the typical thermal stability fuel, the combined effects of copper and accelerated aging, did. Furthermore, the addition of MDA prior to aging of the copper-doped, typical stability fuel significantly counteracted the adverse effect of copper and aging. A similar beneficial effect of MDA was not observed for the poor stability fuel. These results focus on the compositional differences among fuels and the need to elucidate these differences (physical and chemical) for a better understanding and prediction of their performance.

TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

International Nuclear Information System (INIS)

Weicht, U.; Mueller, W.

1983-01-01

1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

Fidelity deviation in quantum teleportation

OpenAIRE

Bang, Jeongho; Ryu, Junghee; Kaszlikowski, Dagomir

2018-01-01

We analyze the performance of quantum teleportation in terms of average fidelity and fidelity deviation. The average fidelity is defined as the average value of the fidelities over all possible input states and the fidelity deviation is their standard deviation, which is referred to as a concept of fluctuation or universality. In the analysis, we find the condition to optimize both measures under a noisy quantum channel---we here consider the so-called Werner channel. To characterize our resu...

Spent fuel transport cask thermal evaluation under normal and accident conditions

Energy Technology Data Exchange (ETDEWEB)

Pugliese, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Lo Frano, R., E-mail: rosa.lofrano@ing.unipi.i [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Forasassi, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy)

2010-06-15

The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the '80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are

High-temperature thermal-chemical analysis of nuclear fuel channels

Energy Technology Data Exchange (ETDEWEB)

Nekhamkin, Y; Rosenband, V; Hasan, D; Elias, E; Wacholder, E; Gany, A [Technion-Israel Inst. of Tech., Haifa (Israel)

1996-12-01

In a severe accident situation, e.g., a postulated loss of coolant accident with a coincident loss of emergency core cooling (LOCA/LOECC), the core may become partially uncovered and steam may become the only coolant available. The thermodynamic conditions in the core, in this case, depend on ability of the steam to effectively remove the fuel decay heat and the heat generated by the exothermic steam/Zircaloy reaction., Therefore, it is important to understand the high-temperature behavior of an oxidizing fuel channel. The main objective of this work is to develop a methodology for calculating the clad temperature and rate of oxidation of a partially covered fuel pin. A criterion is derived to define the importance of the chemical reaction in the overall heat balance. The main parameters affecting the fuel thermal behavior are outlined (authors).

Porous polymer coatings as substrates for the formation of high-fidelity micropatterns by quill-like pens

Directory of Open Access Journals (Sweden)

Michael Hirtz

2013-06-01

Full Text Available We explored the potentials of microarray printing using quill-like microcantilevers onto solid supports that are typically used in microspot printing, including paper, polymeric nitrocellulose and nylon membranes. We compared these membranes with a novel porous poly(2-hydroxyethyl methacrylate-co-ethylene dimethacrylate support (HEMA with narrow pore size distribution in the 150 nm range, which demonstrated advantages in pattern definition, spot homogeneity, and consistent spot delivery of different dyes (phloxine B and bromophenol blue with diameters of several micrometres. The bromophenol blue arrays on HEMA support were used to detect the presence of bovine serum albumin (BSA. In the presence of BSA, the fluorescence spectrum observed from the bromophenol blue microarray exhibited a significant red shift of the maximum emission wavelength. Our results show that the porous HEMA substrates can improve the fidelity and quality of microarrays prepared by using the quill-like microcantilevers. The presented method sets the stage for further studies using chemical and biochemical recognition elements, along with colorimetric and fluorometric sensors that can be spotted by this method onto flat porous polymer substrates.

Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

International Nuclear Information System (INIS)

Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

2013-01-01

A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

Macroscopic behavior of fast reactor fuel subjected to simulated thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Emerson, J.E.; Savoie, F.E.

1983-06-01

High-speed cinematography has been used to characterize the macroscopic behavior of irradiated and unirradiated fuel subjected to thermal transients prototypical of fast reactor transients. The results demonstrate that as the cladding melts, the fuel can disperse via spallation if the fuel contains in excess of approx. 16 μmoles/gm of fission gas. Once the cladding has melted, the macroscopic behavior (time to failure and dispersive nature) was strongly influenced by the presence of volatile fission products and the heating rate

CFD thermal-hydraulic analysis of a CANDU fuel channel

International Nuclear Information System (INIS)

Catana, A.; Prisecaru, I.; Dupleac, D.; Danila, N.

2009-01-01

This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational fluid dynamics) methodology approach. Limited computer power available at Bucharest University POLITEHNICA forced the authors to analyse only segments of fuel channel namely the significant ones: fuel bundle junctions with adjacent segments, fuel bundle spacer planes with adjacent segments, regular segments of fuel bundles. The computer code used is FLUENT. Fuel bundles contained in pressure tubes forms a complex flow domain. The flow is characterized by high turbulence and in some parts of fuel channel also by multi-phase flow. The flow in the fuel channel has been simulated by solving the equations for conservation of mass and momentum. For turbulence modelling the standard k-e model is employed although other turbulence models can be used as well. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Since we consider only some relatively short segments of a CANDU fuel channel we can assume, for this starting stage, that heat transfer is not very important for these short segments of fuel channel. The boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. In this paper we present results for Standard CANDU 6 Fuel Bundles as a basis for CFD thermal-hydraulic analysis of INR proposed SEU43 and other new nuclear fuels. (authors)

On the thermal conductivity of UO2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

International Nuclear Information System (INIS)

Walker, C.T.; Staicu, D.; Sheindlin, M.; Papaioannou, D.; Goll, W.; Sontheimer, F.

2006-01-01

A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation

A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

Directory of Open Access Journals (Sweden)

Jorge Pérez Mañes

2014-01-01

Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

International Nuclear Information System (INIS)

Choi, Hae Yun; Kwon, Jong Soo; Park, Seong Hoon; Kim, Seong Rea; Lee, Gi Won

1996-01-01

The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new

Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Choi, Hae Yun; Kwon, Jong Soo; Park, Seong Hoon; Kim, Seong Rea; Lee, Gi Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-01-01

The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new.

Fully coupled multiphysics modeling of enhanced thermal conductivity UO{sub 2}–BeO fuel performance in a light water reactor

Energy Technology Data Exchange (ETDEWEB)

Liu, R. [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Zhou, W., E-mail: wenzzhou@cityu.edu.hk [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Shen, P. [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Prudil, A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Chan, P.K. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario (Canada)

2015-12-15

Highlights: • LWR fuel performance modeling capability developed. • Fully coupled multiphysics studies for enhanced thermal conductivity UO{sub 2}–BeO fuel. • UO{sub 2}–BeO fuel decreases fuel temperature and lessens thermal stresses. • UO{sub 2}–BeO fuel facilitates a reduction in PCMI. • Reactor safety can be improved for UO{sub 2}–BeO fuel. - Abstract: Commercial light water reactor fuel UO{sub 2} has a low thermal conductivity that leads to the development of a large temperature gradient across the fuel pellet, limiting the reactor operational performance due to the effects that include thermal stresses causing pellet cladding interaction and the release of fission product gases. This study presents the development of a modeling and simulation for enhanced thermal conductivity UO{sub 2}–BeO fuel behavior in a light water reactor, using self-defined multiple physics models fully coupled based on the framework of COMSOL Multiphysics. Almost all the related physical models are considered, including heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the phenomenal models and materials properties are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. UO{sub 2}–BeO enhanced thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction from our simulation results through lessening thermal stresses that result in fuel cracking, relocation, and swelling, so that the safety of the reactor would be improved.

Manufacture of a UO2-Based Nuclear Fuel with Improved Thermal Conductivity with the Addition of BeO

Science.gov (United States)

Garcia, Chad B.; Brito, Ryan A.; Ortega, Luis H.; Malone, James P.; McDeavitt, Sean M.

2017-12-01

The low thermal conductivity of oxide nuclear fuels is a performance-limiting parameter. Enhancing this property may provide a contribution toward establishing accident-tolerant fuel forms. In this study, the thermal conductivity of UO2 was increased through the fabrication of ceramic-ceramic composite forms with UO2 containing a continuous BeO matrix. Fuel with a higher thermal conductivity will have reduced thermal gradients and lower centerline temperatures in the fuel pin. Lower operational temperatures will reduce fission gas release and reduce fuel restructuring. Additions of BeO were made to UO2 fuel pellets in 2.5, 5, 7.5, and 10 vol pct concentrations with the goals of establishing reliable lab-scale processing procedures, minimizing porosity, and maximizing thermal conductivity. The microstructure was characterized with electron probe microanalysis, and the thermal properties were assessed by light flash analysis and differential scanning calorimetry. Reliable, high-density samples were prepared using compaction pressure between 200 and 225 MPa and sintering times between 4 and 6 hours. It was found that the thermal conductivity of UO2 improved approximately 10 pct for each 1 vol pct BeO added over the measured temperature range 298.15 K to 523.15 K (25 °C to 250 °C) with the maximum observed improvement being ˜ 100 pct, or doubled, at 10 vol pct BeO.

Thermal conditions and functional requirements for molten fuel containment

International Nuclear Information System (INIS)

Kang, C.S.; Torri, A.

1980-05-01

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

Uranium-fuel thermal reactor benchmark testing of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2001-01-01

CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

Science.gov (United States)

Andrianova, E. A.; Tsibul'skiy, V. F.

2017-12-01

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

Three-dimensional single-channel thermal analysis of fully ceramic microencapsulated fuel via two-temperature homogenized model

International Nuclear Information System (INIS)

Lee, Yoonhee; Cho, Nam Zin

2014-01-01

Highlights: • Two-temperature homogenized model is applied to thermal analysis of fully ceramic microencapsulated (FCM) fuel. • Based on the results of Monte Carlo calculation, homogenized parameters are obtained. • 2-D FEM/1-D FDM hybrid method for the model is used to obtain 3-D temperature profiles. • The model provides the fuel-kernel and SiC matrix temperatures separately. • Compared to UO 2 fuel, the FCM fuel shows ∼560 K lower maximum temperatures at steady- and transient states. - Abstract: The fully ceramic microencapsulated (FCM) fuel, one of the accident tolerant fuel (ATF) concepts, consists of TRISO particles randomly dispersed in SiC matrix. This high heterogeneity in compositions leads to difficulty in explicit thermal calculation of such a fuel. For thermal analysis of a fuel element of very high temperature reactors (VHTRs) which has a similar configuration to FCM fuel, two-temperature homogenized model was recently proposed by the authors. The model was developed using particle transport Monte Carlo method for heat conduction problems. It gives more realistic temperature profiles, and provides the fuel-kernel and graphite temperatures separately. In this paper, we apply the two-temperature homogenized model to three-dimensional single-channel thermal analysis of the FCM fuel element for steady- and transient-states using 2-D FEM/1-D FDM hybrid method. In the analyses, we assume that the power distribution is uniform in radial direction at steady-state and that in axial direction it is in the form of cosine function for simplicity. As transient scenarios, we consider (i) coolant inlet temperature transient, (ii) inlet mass flow rate transient, and (iii) power transient. The results of analyses are compared to those of conventional UO 2 fuel having the same geometric dimension and operating conditions

Comparative analysis of thermal behavior in hollow nuclear fuel pellets

International Nuclear Information System (INIS)

Santos, Beatriz M. dos; Alvim, Antonio C.M.

2017-01-01

The increase in energy demand in Brazil and in the world is a real problem and several solutions are being considered to mitigate it. Maximization of energy generation, within the safety standards of fuel resources already known, is one of them. In this respect, nuclear energy is a crucial technology to sustain energy demand on several countries. Performances of a solid cylindrical and an annular rod have been verified and compared; where it has been proven that the annular rod can reach a higher nominal power in relation to the solid one. In this paper, the temperature profiles of two distinct nuclear fuel pellets, one of them annular and the other in the shape of a hollow biconcave disc (like the cross section of a red blood cell), were compared to analyze the efficiency and safety of both. The finite differences method allowed the evaluation of the thermal behavior of these pellets, where one specific physical condition was analyzed, regarding convection and conduction at the lateral edges. The results show that the temperature profile of the hollow biconcave disc pellet is lower, about 70 deg C below, when compared to the temperature profile of the annular pellet, considering the same simulation parameters for both pellets. (author)

Comparative analysis of thermal behavior in hollow nuclear fuel pellets

Energy Technology Data Exchange (ETDEWEB)

Santos, Beatriz M. dos; Alvim, Antonio C.M., E-mail: bmachado@nuclear.ufrj.br, E-mail: aalvim@gmail.com [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2017-11-01

The increase in energy demand in Brazil and in the world is a real problem and several solutions are being considered to mitigate it. Maximization of energy generation, within the safety standards of fuel resources already known, is one of them. In this respect, nuclear energy is a crucial technology to sustain energy demand on several countries. Performances of a solid cylindrical and an annular rod have been verified and compared; where it has been proven that the annular rod can reach a higher nominal power in relation to the solid one. In this paper, the temperature profiles of two distinct nuclear fuel pellets, one of them annular and the other in the shape of a hollow biconcave disc (like the cross section of a red blood cell), were compared to analyze the efficiency and safety of both. The finite differences method allowed the evaluation of the thermal behavior of these pellets, where one specific physical condition was analyzed, regarding convection and conduction at the lateral edges. The results show that the temperature profile of the hollow biconcave disc pellet is lower, about 70 deg C below, when compared to the temperature profile of the annular pellet, considering the same simulation parameters for both pellets. (author)

Numerical Simulations of High Reactivity Gasoline Fuel Sprays under Vaporizing and Reactive Conditions

KAUST Repository

Mohan, Balaji; Jaasim, Mohammed; Ahmed, Ahfaz; Hernandez Perez, Francisco; Sim, Jaeheon; Roberts, William L.; Sarathy, Mani; Im, Hong G.

2018-01-01

Gasoline compression ignition (GCI) engines are becoming more popular alternative for conventional spark engines to harvest the advantage of high volatility. Recent experimental study demonstrated that high reactivity gasoline fuel can be operated in a conventional mixing controlled combustion mode producing lower soot emissions than that of diesel fuel under similar efficiency and NOx level [1]. Therefore, there is much interest in using gasoline-like fuels in compression ignition engines. In order to improve the fidelity of simulation-based GCI combustion system development, it is mandatory to enhance the prediction of spray combustion of gasoline-like fuels. The purpose of this study is to model the spray characteristics of high reactivity gasoline fuels and validate the models with experimental results obtained through an optically accessible constant volume vessel under vaporizing [2] and reactive conditions [3]. For reacting cases, a comparison of PRF and KAUST multi-component surrogate (KMCS) mechanism was done to obtain good agreement with the experimental ignition delay. From this study, some recommendations were proposed for GCI combustion modelling framework using gasoline like fuels.

Numerical Simulations of High Reactivity Gasoline Fuel Sprays under Vaporizing and Reactive Conditions

KAUST Repository

Mohan, Balaji

2018-04-03

Gasoline compression ignition (GCI) engines are becoming more popular alternative for conventional spark engines to harvest the advantage of high volatility. Recent experimental study demonstrated that high reactivity gasoline fuel can be operated in a conventional mixing controlled combustion mode producing lower soot emissions than that of diesel fuel under similar efficiency and NOx level [1]. Therefore, there is much interest in using gasoline-like fuels in compression ignition engines. In order to improve the fidelity of simulation-based GCI combustion system development, it is mandatory to enhance the prediction of spray combustion of gasoline-like fuels. The purpose of this study is to model the spray characteristics of high reactivity gasoline fuels and validate the models with experimental results obtained through an optically accessible constant volume vessel under vaporizing [2] and reactive conditions [3]. For reacting cases, a comparison of PRF and KAUST multi-component surrogate (KMCS) mechanism was done to obtain good agreement with the experimental ignition delay. From this study, some recommendations were proposed for GCI combustion modelling framework using gasoline like fuels.

Sensitivity Analysis of Heavy Fuel Oil Spray and Combustion under Low-Speed Marine Engine-Like Conditions

Directory of Open Access Journals (Sweden)

Lei Zhou

2017-08-01

Full Text Available On account of their high power, thermal efficiency, good reliability, safety, and durability, low-speed two-stroke marine diesel engines are used as the main drive devices for large fuel and cargo ships. Most marine engines use heavy fuel oil (HFO as the primary fuel, however, the physical and chemical characteristics of HFO are not clear because of its complex thermophysical properties. The present study was conducted to investigate the effects of fuel properties on the spray and combustion characteristics under two-stroke marine engine-like conditions via a sensitivity analysis. The sensitivity analysis of fuel properties for non-reacting and reacting simulations are conducted by comparing two fuels having different physical properties, such as fuel density, dynamic viscosity, critical temperature, and surface tension. The performances of the fuels are comprehensively studied under different ambient pressures, ambient temperatures, fuel temperatures, and swirl flow conditions. From the results of non-reacting simulations of HFO and diesel fuel properties in a constant volume combustion chamber, it can be found that the increase of the ambient pressure promotes fuel evaporation, resulting in a reduction in the steady liquid penetration of both diesel and HFO; however, the difference in the vapor penetrations of HFO and diesel reduces. Increasing the swirl flow significantly influences the atomization of both HFO and diesel, especially the liquid distribution of diesel. It is also found that the ambient temperature and fuel temperature have the negative effects on Sauter mean diameter (SMD distribution. For low-speed marine engines, the combustion performance of HFO is not sensitive to activation energy in a certain range of activation energy. At higher engine speed, the difference in the effects of different activation energies on the in-cylinder pressure increases. The swirl flow in the cylinder can significantly promote fuel evaporation and

Efficient Biomass Fuel Cell Powered by Sugar with Photo- and Thermal-Catalysis by Solar Irradiation.

Science.gov (United States)

Liu, Wei; Gong, Yutao; Wu, Weibing; Yang, Weisheng; Liu, Congmin; Deng, Yulin; Chao, Zi-Sheng

2018-06-19

The utilization of biomass sugars has received great interesting recently. Herein, we present a highly efficient hybrid solar biomass fuel cell that utilizes thermal- and photocatalysis of solar irradiation and converts biomass sugars into electricity with high power output. The fuel cell uses polyoxometalates (POMs) as photocatalyst to decompose sugars and capture their electrons. The reduced POMs have strong visible and near-infrared light adsorption, which can significantly increase the temperature of the reaction system and largely promotes the thermal oxidation of sugars by the POM. In addition, the reduced POM functions as charge carrier that can release electrons at the anode in the fuel cell to generate electricity. The electron-transfer rates from glucose to POM under thermal and light-irradiation conditions were investigated in detail. The power outputs of this solar biomass fuel cell are investigated by using different types of sugars as fuels, with the highest power density reaching 45 mW cm -2 . © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Optimization of thermal efficiency of nuclear central power like as PWR

International Nuclear Information System (INIS)

Lapa, Nelbia da Silva

2005-10-01

The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

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

International Nuclear Information System (INIS)

Verbeek, P.; Hoppe, N.

1976-01-01

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

THERMAL PHASES OF EARTH-LIKE PLANETS: ESTIMATING THERMAL INERTIA FROM ECCENTRICITY, OBLIQUITY, AND DIURNAL FORCING

International Nuclear Information System (INIS)

Cowan, Nicolas B.; Voigt, Aiko; Abbot, Dorian S.

2012-01-01

In order to understand the climate on terrestrial planets orbiting nearby Sun-like stars, one would like to know their thermal inertia. We use a global climate model to simulate the thermal phase variations of Earth analogs and test whether these data could distinguish between planets with different heat storage and heat transport characteristics. In particular, we consider a temperate climate with polar ice caps (like the modern Earth) and a snowball state where the oceans are globally covered in ice. We first quantitatively study the periodic radiative forcing from, and climatic response to, rotation, obliquity, and eccentricity. Orbital eccentricity and seasonal changes in albedo cause variations in the global-mean absorbed flux. The responses of the two climates to these global seasons indicate that the temperate planet has 3× the bulk heat capacity of the snowball planet due to the presence of liquid water oceans. The obliquity seasons in the temperate simulation are weaker than one would expect based on thermal inertia alone; this is due to cross-equatorial oceanic and atmospheric energy transport. Thermal inertia and cross-equatorial heat transport have qualitatively different effects on obliquity seasons, insofar as heat transport tends to reduce seasonal amplitude without inducing a phase lag. For an Earth-like planet, however, this effect is masked by the mixing of signals from low thermal inertia regions (sea ice and land) with that from high thermal inertia regions (oceans), which also produces a damped response with small phase lag. We then simulate thermal light curves as they would appear to a high-contrast imaging mission (TPF-I/Darwin). In order of importance to the present simulations, which use modern-Earth orbital parameters, the three drivers of thermal phase variations are (1) obliquity seasons, (2) diurnal cycle, and (3) global seasons. Obliquity seasons are the dominant source of phase variations for most viewing angles. A pole-on observer

THERMAL PHASES OF EARTH-LIKE PLANETS: ESTIMATING THERMAL INERTIA FROM ECCENTRICITY, OBLIQUITY, AND DIURNAL FORCING

Energy Technology Data Exchange (ETDEWEB)

Cowan, Nicolas B. [Center for Interdisciplinary Exploration and Research in Astrophysics and Department of Physics and Astronomy, Northwestern University, 2131 Tech Drive, Evanston, IL 60208 (United States); Voigt, Aiko [Max Planck Institute for Meteorology, Bundesstr. 53, D-20146 Hamburg (Germany); Abbot, Dorian S., E-mail: n-cowan@nortwestern.edu [Department of Geophysical Sciences, University of Chicago, 5734 South Ellis Avenue, Chicago, IL 60637 (United States)

2012-09-20

In order to understand the climate on terrestrial planets orbiting nearby Sun-like stars, one would like to know their thermal inertia. We use a global climate model to simulate the thermal phase variations of Earth analogs and test whether these data could distinguish between planets with different heat storage and heat transport characteristics. In particular, we consider a temperate climate with polar ice caps (like the modern Earth) and a snowball state where the oceans are globally covered in ice. We first quantitatively study the periodic radiative forcing from, and climatic response to, rotation, obliquity, and eccentricity. Orbital eccentricity and seasonal changes in albedo cause variations in the global-mean absorbed flux. The responses of the two climates to these global seasons indicate that the temperate planet has 3 Multiplication-Sign the bulk heat capacity of the snowball planet due to the presence of liquid water oceans. The obliquity seasons in the temperate simulation are weaker than one would expect based on thermal inertia alone; this is due to cross-equatorial oceanic and atmospheric energy transport. Thermal inertia and cross-equatorial heat transport have qualitatively different effects on obliquity seasons, insofar as heat transport tends to reduce seasonal amplitude without inducing a phase lag. For an Earth-like planet, however, this effect is masked by the mixing of signals from low thermal inertia regions (sea ice and land) with that from high thermal inertia regions (oceans), which also produces a damped response with small phase lag. We then simulate thermal light curves as they would appear to a high-contrast imaging mission (TPF-I/Darwin). In order of importance to the present simulations, which use modern-Earth orbital parameters, the three drivers of thermal phase variations are (1) obliquity seasons, (2) diurnal cycle, and (3) global seasons. Obliquity seasons are the dominant source of phase variations for most viewing angles. A

Thermal design and analysis of the HTGR fuel element vertical carbonizing and annealing furnace

International Nuclear Information System (INIS)

Llewellyn, G.H.

1977-06-01

Computer analyses of the thermal design for the proposed HTGR fuel element vertical carbonizing and annealing furnace were performed to verify its capability and to determine the required power input and distribution. Although the furnace is designed for continuous operation, steady-state temperature distributions were obtained by assuming internal heat generation in the fuel elements to simulate their mass movement. The furnace thermal design, the analysis methods, and the results are discussed herein

A Novel Low Temperature PCR Assured High-Fidelity DNA Amplification

Directory of Open Access Journals (Sweden)

Shaoxia Zhou

2013-06-01

Full Text Available As previously reported, a novel low temperature (LoTemp polymerase chain reaction (PCR catalyzed by a moderately heat-resistant (MHR DNA polymerase with a chemical-assisted denaturation temperature set at 85 °C instead of the conventional 94–96 °C can achieve high-fidelity DNA amplification of a target DNA, even after up to 120 PCR thermal cycles. Furthermore, such accurate amplification is not achievable with conventional PCR. Now, using a well-recognized L1 gene segment of the human papillomavirus (HPV type 52 (HPV-52 as the template for experiments, we demonstrate that the LoTemp high-fidelity DNA amplification is attributed to an unusually high processivity and stability of the MHR DNA polymerase whose high fidelity in template-directed DNA synthesis is independent of non-existent 3'–5' exonuclease activity. Further studies and understanding of the characteristics of the LoTemp PCR technology may facilitate implementation of DNA sequencing-based diagnostics at the point of care in community hospital laboratories.

Spent fuel pool thermal-hydraulic analysis using RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)

Thermal analysis for a spent reactor fuel storage test in granite

International Nuclear Information System (INIS)

Montan, D.N.

1980-09-01

A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m 3 /s and an ambient rock temperature of 23 0 C, the maximum calculated rock temperature (near the center of the heat source) is about 100 0 C while the maximum air temperature in the drift is around 40 0 C. This ventilation (1 m 3 /s through the main drift and 1/2 m 3 /s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage

COBRA-SFS thermal analysis of a sealed storage cask for the Monitored Retrievable Storage of spent fuel

International Nuclear Information System (INIS)

Rector, D.R.; Wheeler, C.L.

1986-01-01

The COBRA-SFS (Spent Fuel Storage) computer code was used to predict temperature distributions in a concrete Sealed Storage Cask (SSC). This cask was designed for the Department of Energy in the Monitored Retrievable Storage (MRS) program for storage of spent fuel from commercial power operations. Analytical results were obtained for nominal operation of the SSC with spent fuel from 36 PWR fuel assemblies consolidated in 12 cylindrical canisters. Each canister generates 1650 W of thermal power. A parametric study was performed to assess the effects on cask thermal performance of thermal conductivity of the concrete, the fin material, and the amount of radial reinforcing steel bars (rebar). Seven different cases were modeled. The results of the COBRA-SFS analysis of the current cask design predict that the peak fuel cladding temperature in the SSC will not exceed the 37 0 C design limit for the maximum spent fuel load of 19.8 kW and a maximum expected ambient temperature of 37.8 0 C (100 0 F). The results of the parametric analyses illustrate the importance of material selection and design optimization with regard to the SSC thermal performance

Early-in-life thermal performance of UO2--PuO2 fast reactor fuel

International Nuclear Information System (INIS)

Baker, R.B.; Leggett, R.D.

1979-01-01

Results from the combined analyses of two thermal performance tests, HEDL P-19 and HEDL P-20 are described. The tests were designed to provide data on the power required to cause incipient fuel melting early in life under conditions prototypic of FFTF driver fuel pins and similar FBR fuel systems

Combustion and emission characteristics of diesel engine fueled with diesel-like fuel from waste lubrication oil

International Nuclear Information System (INIS)

Wang, Xiangli; Ni, Peiyong

2017-01-01

Highlights: • 100% diesel-like fuel from waste lubricating oil was conducted in a diesel engine. • Good combustion and fuel economy are achieved without engine modifications. • Combustion duration of DLF is shorter than diesel. • NOx and smoke emissions with the DLF are slightly higher than pure diesel. - Abstract: Waste lubricant oil (WLO) is one of the most important types of the energy sources. WLO cannot be burned directly in diesel engines, but can be processed to be used as diesel-like fuel (DLF) to minimize its harmful effect and maximize its useful values. Moreover, there are some differences in physicochemical properties between WLO and diesel fuel. In order to identify the differences in combustion and emission performance of diesel engine fueled with the two fuels, a bench test of a single-cylinder direct injection diesel engine without any engine modification was investigated at four engine speeds and five engine loads. The effects of the fuels on fuel economic performance, combustion characteristics, and emissions of hydrocarbons (HC), carbon monoxide (CO), nitrogen oxides (NOx) and smoke were discussed. The DLF exhibits longer ignition delay period and shorter combustion duration than diesel fuel. The test results indicate that the higher distillation temperatures of the DLF attribute to the increase of combustion pressure, temperature and heat release rate. The brake specific fuel consumption (BSFC) of the DLF compared to diesel is reduced by about 3% at 3000 rpm under light and medium loads. The DLF produces slightly higher NOx emissions at middle and heavy loads, somewhat more smoke emissions at middle loads, and notably higher HC and CO emissions at most measured points than diesel fuel. It is concluded that the DLF can be used as potential available fuel in high-speed diesel engines without any problems.

Enhancing experience prototyping by the help of mixed-fidelity prototypes

OpenAIRE

Yasar, Ansar-Ul-Haque

2007-01-01

In this research review I undertook the problem related to the usage of a new concept known as the Mixed- Fidelity Prototype which is a mixture of its predecessors Low- and High- Fidelity Prototypes in Experience Prototyping. Experience Prototyping is a good way to explore, communicate and interact with the designs we develop like experiencing cycling on the ice, although the mood, snow conditions, bicycle type and many other factors really matter and tend to change with time. Experience Prot...

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

Science.gov (United States)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Advanced three-dimensional thermal modeling of a baseline spent fuel repository

International Nuclear Information System (INIS)

Altenbach, T.J.; Lowry, W.E.

1980-01-01

A three-dimensional thermal analysis using finite difference techniques was performed to determine the near-field response of a baseline spent fuel repository in a deep geologic salt medium. A baseline design incorporates previous thermal modeling experience and OWI recommendations for areal thermal loading in specifying the waste form properties, package details, and emplacement configuration. The base case in this thermal analysis considers one 10-year old PWR spent fuel assembly emplaced to yield a 36 kW/acre (8.9 W/m 2 ) loading. A unit cell model in an infinite array is used to simplify the problem and provide upper-bound temperatures. Boundary conditions are imposed which allow simulations to 1000 years. Variations studied include a comparison of ventilated and unventilated storage room conditions, emplacement packages with and without air gaps surrounding the canister, and room cool-down scenarios with ventilation following an unventilated state for retrieval purposes. It was found that at this low-power level, ventilating the emplacement room has an immediate cooling influence on the canister and effectively maintains the emplacement room floor near the temperature of the ventilating air

FIBWR2 evaluation of fuel thermal limits during density wave oscillaions in BWRs

Energy Technology Data Exchange (ETDEWEB)

Nik, N.; Rajan, S.R.; Karasulu, M. [New York Power Authority, White Plains, NY (United States)

1995-09-01

Analyses were performed to evaluate hydraulic and thermal margin responses of three different BWR fuel designs subjected to the same periodic power/flow oscillations, such as those that might be exhibited during an instability event. The power/flow versus time information from the oscillations was used as a forcing function to calculate the hydraulic response and the MCPR performance of the limiting fuel bundles during the regional oscillations using the analytical code FIBWR2. The results of the calculations were used to determine the thermal margin variation as a function of oscillation magnitude.

Thermal modeling and temperature control of a PEM fuel cell system for forklift applications

DEFF Research Database (Denmark)

Liso, Vincenzo; Nielsen, Mads Pagh; Kær, Søren Knudsen

2014-01-01

Temperature changes in PEM fuel cell stacks are considerably higher during load variations and have a negative impact as they generate thermal stresses and stack degradation. Cell hydration is also of vital importance in fuel cells and it is strongly dependent on operating temperature....... A combination of high temperature and reduced humidity increases the degradation rate. Stack thermal management and control are, thus, crucial issues in PEM fuel cell systems especially in automotive applications such as forklifts. In this paper we present a control–oriented dynamic model of a liquid–cooled PEM...... fuel cell system for studying temperature variations over fast load changes. A temperature dependent cell polarization and hydration model integrated with the compressor, humidifier and cooling system are simulated in dynamic condition. A feedback PID control was implemented for stack cooling...

Thermal Analysis of Cold Vacuum Drying (CVD) of Spent Nuclear Fuel (SNF)

International Nuclear Information System (INIS)

PIEPHO, M.G.

2000-01-01

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 N Reactor spent fuel. This analysis provides the basis for the MCO thermal behavior at the CVD Facility in support of the safety basis documentation

Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

International Nuclear Information System (INIS)

Heard, F.J.; Cramer, E.R.; Beaver, T.R.; Thurgood, M.J.

1996-01-01

A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy's Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses

Thermally decomposed ricebran oil as a diesel fuel

Directory of Open Access Journals (Sweden)

Megahed, O. A.

1998-04-01

Full Text Available Ricebran oil; a non edible oil, was thermally decomposed using different loads of calcium oxide as catalyst. The fuel properties of the cracked product were evaluated as compared to those of diesel fuel. The considered properties included the calorific value, flash point, viscosity, pour point, distillation characteristics, cetane number in addition to some other fuel properties. The results had shown that the fuel properties of the decomposed oil were quite similar to those of standard diesel fuel. The calorific value was 80-90% that of diesel fuel and the viscosity was sligthy higher. The prepared fuel was advantageous over diesel fuel as the former was completely free from sulfur, which on fuel combustion produces corrosive gases of sulfur oxides.

Aceite de germen de arroz, un aceite no comestible, fue descompuesto térmicamente usando diferentes cantidades de óxido cálcico como catalizador. Las propiedades combustibles del producto craqueado fueron evaluadas comparándolas con las del gasóleo. Las propiedades consideradas incluyeron el poder calorífico, punto de inflamación, viscosidad, temperatura de fluidez crítica, características de destilación, número de cetano y otras propiedades de los combustibles. Los resultados han mostrado que las propiedades combustibles del aceite descompuesto fueron bastantes similares a la de los gasóleos estándar. El poder calorífico fue del 80-90% de la del gasóleo y la viscosidad ligeramente mayor. El combustible preparado fue ventajoso sobre el gasóleo ya que el primero estaba completamente libre de sulfuro, el cual produce en la combustión del carburante gases corrosivos de óxido de azufre.

Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

International Nuclear Information System (INIS)

Pantoja C, R.

2010-01-01

To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

Spent-fuel pool thermal hydraulics: The evaporation question

International Nuclear Information System (INIS)

Yilmaz, T.P.; Lai, J.C.

1996-01-01

Many nuclear power plants are currently using dense fuel arrangements that increase the number of spent fuel elements stored in their spent-fuel pools (SFPs). The denser spent-fuel storage results in higher water temperatures, especially when certain event scenarios are analyzed. In some of these event scenarios, it is conservative to maximize the evaporation rate, while in other circumstances it is required to minimize the evaporation rates for conservatism. Evaporation is such a fundamental phenomenon that many branches of engineering developed various equations based on theory and experiments. The evaporation rates predicted by existing equations present a wide range of variation, especially at water temperatures >40 degrees C. Furthermore, a study on which equations provide the highest and lowest evaporation rates has not been done until now. This study explores the sensitivity of existing evaporation equations to various parameters and recommends the limiting evaporation equations for use in the solution of SFP thermal problems. Note that the results of this study may be applicable to a much wider range of applications from irrigation ponds, cooling lakes, and liquid-waste management to calculating adequate air exchange rate for swimming pools and health spas

Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

International Nuclear Information System (INIS)

Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

2002-12-01

In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

Design of an optical thermal sensor for proton exchange membrane fuel cell temperature measurement using phosphor thermometry

Science.gov (United States)

Inman, Kristopher; Wang, Xia; Sangeorzan, Brian

Internal temperatures in a proton exchange membrane (PEM) fuel cell govern the ionic conductivities of the polymer electrolyte, influence the reaction rate at the electrodes, and control the water vapor pressure inside the cell. It is vital to fully understand thermal behavior in a PEM fuel cell if performance and durability are to be optimized. The objective of this research was to design, construct, and implement thermal sensors based on the principles of the lifetime-decay method of phosphor thermometry to measure temperatures inside a PEM fuel cell. Five sensors were designed and calibrated with a maximum uncertainty of ±0.6 °C. Using these sensors, surface temperatures were measured on the cathode gas diffusion layer of a 25 cm 2 PEM fuel cell. The test results demonstrate the utility of the optical temperature sensor design and provide insight into the thermal behavior found in a PEM fuel cell.

Transportation fuel production by combination of LDPE thermal cracking and catalytic hydroreforming.

Science.gov (United States)

Escola, J M; Aguado, J; Serrano, D P; Briones, L

2014-11-01

Fuel production from plastics is a promising way to reduce landfilling rates while obtaining valuable products. The usage of Ni-supported hierarchical Beta zeolite (h-Beta) for the hydroreforming of the oils coming from LDPE thermal cracking has proved to produce high selectivities to gasoline and diesel fuels (>80%). In the present work, the effect of the Ni loading on Ni/h-Beta is investigated in the hydroreforming of the oils form LDPE thermal cracking. h-Beta samples were impregnated with Ni nitrate, calcined and reduced in H2 up to 550°C to achieve different Ni contents: 1.5%, 4%, 7% and 10%. Larger and more easily reducible metal particles were obtained on Ni 7%/h-Beta and Ni 10%/h-Beta. Hydroreforming tests were carried out in autoclave reactor at 310°C, under 20 bar H2, for 45 min. Ni content progressively increased the amount of gases at the expenses of diesel fractions, while gasoline remained approximately constant about 52-54%. Maximum selectivity to automotive fuels (∼81%) was obtained with Ni 7%/h-Beta. Ni loading also enhanced olefins saturation up to Ni 7%/h-Beta. High cetane indices (71-86) and octane numbers (89-91) were obtained over all the catalysts. Regarding the different studied Ni contents, Ni 7%/h-Beta constitutes a rather promising catalyst for obtaining high quality fuels from LDPE thermal cracking oils. Copyright © 2014 Elsevier Ltd. All rights reserved.

Gas Test Loop Booster Fuel Hydraulic Testing

International Nuclear Information System (INIS)

Gas Test Loop Hydraulic Testing Staff

2006-01-01

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

Gas Test Loop Booster Fuel Hydraulic Testing

Energy Technology Data Exchange (ETDEWEB)

Gas Test Loop Hydraulic Testing Staff

2006-09-01

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

Fidelity Failures in Brief Strategic Family Therapy for Adolescent Drug Abuse: A Clinical Analysis.

Science.gov (United States)

Lebensohn-Chialvo, Florencia; Rohrbaugh, Michael J; Hasler, Brant P

2018-04-30

As evidence-based family treatments for adolescent substance use and conduct problems gain traction, cutting edge research moves beyond randomized efficacy trials to address questions such as how these treatments work and how best to disseminate them to community settings. A key factor in effective dissemination is treatment fidelity, which refers to implementing an intervention in a manner consistent with an established manual. While most fidelity research is quantitative, this study offers a qualitative clinical analysis of fidelity failures in a large, multisite effectiveness trial of Brief Strategic Family Therapy (BSFT) for adolescent drug abuse, where BSFT developers trained community therapists to administer this intervention in their own agencies. Using case notes and video recordings of therapy sessions, an independent expert panel first rated 103 cases on quantitative fidelity scales grounded in the BSFT manual and the broader structural-strategic framework that informs BSFT intervention. Because fidelity was generally low, the panel reviewed all cases qualitatively to identify emergent types or categories of fidelity failure. Ten categories of failures emerged, characterized by therapist omissions (e.g., failure to engage key family members, failure to think in threes) and commissions (e.g., off-model, nonsystemic formulations/interventions). Of these, "failure to think in threes" appeared basic and particularly problematic, reflecting the central place of this idea in structural theory and therapy. Although subject to possible bias, our observations highlight likely stumbling blocks in exporting a complex family treatment like BSFT to community settings. These findings also underscore the importance of treatment fidelity in family therapy research. © 2018 Family Process Institute.

Retrofitting of homes for people in fuel poverty: Approach based on household thermal comfort

International Nuclear Information System (INIS)

Vilches, Alberto; Barrios Padura, Ángela; Molina Huelva, Marta

2017-01-01

In a context of fuel poverty, the application of cost-effective methodology for energy retrofitting of buildings is ineffective. In these situations, there is no energy consumption reduction and thermal comfort is hardly achieved. This paper introduces a methodology to choose the most appropriate retrofit measure in a context of fuel poverty. This methodology is based on thermal comfort as the main criterion, and secondarily, it is based on the budget for paying monthly energy bills and initial costs. This study demonstrates how thermal comfort and monthly available income of households must be the first criterion for decision-making process. The methodology has been applied in four buildings. Results demonstrate why active retrofit measures are ineffective when monthly budget to pay energy bills is reduced. In conclusion, recommendations are made to promote efficient public policies in energy retrofit. The study was supported by the European Regional Development Fund (ERDF) and carried out within the research project ‘{Re} Programa. (Re)habitation +(Re)generation +(Re)programming’ during 2013–2015. - Highlights: • A methodology is proposed for energy retrofit in a fuel poverty context. • Householder's monthly budget determines the energy retrofit effectiveness. • The cost-effective methodology is not applicable in a fuel poverty context. • Energy retrofit does not reduce energy consumption in a fuel poverty context. • Energy retrofit does not produce monthly monetary savings in a fuel poverty context.

Innovative nuclear thermal rocket concept utilizing LEU fuel for space application

International Nuclear Information System (INIS)

Nam, Seung Hyun; Venneri, Paolo; Choi, Jae Young; Jeong, Yong Hoon; Chang, Soon Heung

2015-01-01

Space is one of the best places for humanity to turn to keep learning and exploiting. A Nuclear Thermal Rocket (NTR) is a viable and more efficient option for human space exploration than the existing Chemical Rockets (CRs) which are highly inefficient for long-term manned missions such as to Mars and its satellites. NERVA derived NTR engines have been studied for the human missions as a mainstream in the United States of America (USA). Actually, the NERVA technology has already been developed and successfully tested since 1950s. The state-of-the-art technology is based on a Hydrogen gas (H_2) cooled high temperature reactor with solid core utilizing High-Enriched Uranium (HEU) fuel to reduce heavy metal mass and to use fast or epithermal neutron spectrums enabling simple core designs. However, even though the NTR designs utilizing HEU is the best option in terms of rocket performance, they inevitably provoke nuclear proliferation obstacles on all Research and Development (R and D) activities by civilians and non-nuclear weapon states, and its eventual commercialization. To surmount the security issue to use HEU fuel for a NTR, a concept of the innovative NTR engine, Korea Advanced NUclear Thermal Engine Rocket utilizing Low-Enriched Uranium fuel (KANUTER-LEU) is presented in this paper. The design goal of KANUTER-LEU is to make use of a LEU fuel for its compact reactor, but does not sacrifice the rocket performance relative to the traditional NTRs utilizing HEU. KANUTER-LEU mainly consists of a fission reactor utilizing H_2 propellant, a propulsion system and an optional Electricity Generation System as a bimodal engine. To implement LEU fuel for the reactor, the innovative engine adopts W-UO_2 CERMET fuel to drastically increase uranium density and thermal neutron spectrum to improve neutron economy in the core. The moderator and structural material selections also consider neutronic and thermo-physical characteristics to reduce non-fission neutron loss and

Thermal hydraulics of accelerator breeder systems for regeneration of reactor fuel assemblies

International Nuclear Information System (INIS)

Yu, W.S.; Powell, J.R.

1979-01-01

The following conclusions are obtained with regard to the thermal-hydraulic behavior of the Linear Accelerator Fuel Regenerator for PWR and CANDU fuel: (1) two-phase flow is a feasible coolant option for fuel element heat fluxes up to 1 x PWR (or CANDU) average value, which is the maximum design value for a LAFR; (2) two-phase flow pressure drops are low (typically 10 to 30 psi) and film temperature drops very low (typically approx. 10 0 F) for PWR fuel, with inlet velocity range (50 to 75 ft/sec). A somewhat higher inlet velocity range (75 to 100 ft/sec) and pressure drop (50 to 100 psi) is necessary for CANDU fuel, however, to prevent dry out

Investigating variations in implementation fidelity of an organizational-level occupational health intervention.

Science.gov (United States)

Augustsson, Hanna; von Thiele Schwarz, Ulrica; Stenfors-Hayes, Terese; Hasson, Henna

2015-06-01

The workplace has been suggested as an important arena for health promotion, but little is known about how the organizational setting influences the implementation of interventions. The aims of this study are to evaluate implementation fidelity in an organizational-level occupational health intervention and to investigate possible explanations for variations in fidelity between intervention units. The intervention consisted of an integration of health promotion, occupational health and safety, and a system for continuous improvements (Kaizen) and was conducted in a quasi-experimental design at a Swedish hospital. Implementation fidelity was evaluated with the Conceptual Framework for Implementation Fidelity and implementation factors used to investigate variations in fidelity with the Framework for Evaluating Organizational-level Interventions. A multi-method approach including interviews, Kaizen notes, and questionnaires was applied. Implementation fidelity differed between units even though the intervention was introduced and supported in the same way. Important differences in all elements proposed in the model for evaluating organizational-level interventions, i.e., context, intervention, and mental models, were found to explain the differences in fidelity. Implementation strategies may need to be adapted depending on the local context. Implementation fidelity, as well as pre-intervention implementation elements, is likely to affect the implementation success and needs to be assessed in intervention research. The high variation in fidelity across the units indicates the need for adjustments to the type of designs used to assess the effects of interventions. Thus, rather than using designs that aim to control variation, it may be necessary to use those that aim at exploring and explaining variation, such as adapted study designs.

Fuel management service for Tarapur Atomic Power Station core thermal hydraulics

International Nuclear Information System (INIS)

Saha, D.; Venkat Raj, V.; Markandeya, S.G.

1977-01-01

Core thermal hydraulic analysis forms an integral part of the fuel management service for the Tarapur reactors. A distinguishing feature of boiling water reactors is the dependence of core flow distribution on the power distribution. Because of the changes in the axial and radial power distribution from cycle to cycle as well as during the cycle and also the variations in leakage flow, it is necessary to evaluate the core thermal hydraulic parameters for every cycle. Some of the typical results obtained in the course of analysis for different cycles of both the units at Tarapur are presented. The use of MCPR (Minimum Critical Power Ratio), instead of MCHFR (Minimum Critical Heat Flux Ratio) as a figure of merit for fuel cladding integrity is also discussed. (K.B.)

Thermal and stress analysis of a fuel rod research project 277

International Nuclear Information System (INIS)

1975-04-01

The purpose of this investigation was to perform an analytical evaluation of a postulated loss of coolant incident in a large pressurized water reactor. A coupled thermal and stress finite element analysis of a fuel rod subjected to a hypothetical blow-down transient was carried out. The effect of two gap conditions and two initial stress states on the response of the fuel rod was studied. Both one-dimensional and three-dimensional models were investigated. To study the heat transfer in the gap region one assumes a conductive mode of heat transfer in the gap characterized by an equivalent thermal conductivity, which is dependent on the current gap width. Accordingly, coupled analysis procedure and computational scheme were established. A mesh generation computer program was developed for the three-dimensional model

Uncertainties in modeling and scaling in the prediction of fuel stored energy and thermal response

International Nuclear Information System (INIS)

Wulff, W.

1987-01-01

The steady-state temperature distribution and the stored energy in nuclear fuel elements are computed by analytical methods and used to rank, in the order of importance, the effects on stored energy from statistical uncertainties in modeling parameters, in boundary and in operating conditions. An integral technique is used to calculate the transient fuel temperature and to estimate the uncertainties in predicting the fuel thermal response and the peak clad temperature during a large-break loss of coolant accident. The uncertainty analysis presented here is an important part of evaluating the applicability, the uncertainties and the scaling capabilities of computer codes for nuclear reactor safety analyses. The methods employed in this analysis merit general attention because of their simplicity. It is shown that the blowdown peak is dominated by fuel stored energy alone or, equivalently, by linear heating rate. Gap conductance, peaking factors and fuel thermal conductivity are the three most important fuel modeling parameters affecting peak clad temperature uncertainty. 26 refs., 10 figs., 6 tabs

Carbon nanotube-copper exhibiting metal-like thermal conductivity and silicon-like thermal expansion for efficient cooling of electronics.

Science.gov (United States)

Subramaniam, Chandramouli; Yasuda, Yuzuri; Takeya, Satoshi; Ata, Seisuke; Nishizawa, Ayumi; Futaba, Don; Yamada, Takeo; Hata, Kenji

2014-03-07

Increasing functional complexity and dimensional compactness of electronic devices have led to progressively higher power dissipation, mainly in the form of heat. Overheating of semiconductor-based electronics has been the primary reason for their failure. Such failures originate at the interface of the heat sink (commonly Cu and Al) and the substrate (silicon) due to the large mismatch in thermal expansion coefficients (∼300%) of metals and silicon. Therefore, the effective cooling of such electronics demands a material with both high thermal conductivity and a similar coefficient of thermal expansion (CTE) to silicon. Addressing this demand, we have developed a carbon nanotube-copper (CNT-Cu) composite with high metallic thermal conductivity (395 W m(-1) K(-1)) and a low, silicon-like CTE (5.0 ppm K(-1)). The thermal conductivity was identical to that of Cu (400 W m(-1) K(-1)) and higher than those of most metals (Ti, Al, Au). Importantly, the CTE mismatch between CNT-Cu and silicon was only ∼10%, meaning an excellent compatibility. The seamless integration of CNTs and Cu was achieved through a unique two-stage electrodeposition approach to create an extensive and continuous interface between the Cu and CNTs. This allowed for thermal contributions from both Cu and CNTs, resulting in high thermal conductivity. Simultaneously, the high volume fraction of CNTs balanced the thermal expansion of Cu, accounting for the low CTE of the CNT-Cu composite. The experimental observations were in good quantitative concurrence with the theoretically described 'matrix-bubble' model. Further, we demonstrated identical in-situ thermal strain behaviour of the CNT-Cu composite to Si-based dielectrics, thereby generating the least interfacial thermal strain. This unique combination of properties places CNT-Cu as an isolated spot in an Ashby map of thermal conductivity and CTE. Finally, the CNT-Cu composite exhibited the greatest stability to temperature as indicated by its low

Site fidelity, mate fidelity, and breeding dispersal in American kestrels

Science.gov (United States)

Steenhof, K.; Peterson, B.E.

2009-01-01

We assessed mate fidelity, nest-box fidelity, and breeding dispersal distances of American Kestrels (falco sparverius) nesting in boxes in southwestern Idaho from 1990 through 2006. Seventy-seven percent of boxes had different males and 87% had different females where nest-box occupants were identified in consecutive years. High turnover rates were partly a result of box-switching. Forty-eight percent of males and 58% of females that nested within the study area in successive years used different boxes. The probability of changing boxes was unrelated to gender, nesting success in the prior year, or years of nesting experience. Breeding dispersal distances for birds that moved to different boxes averaged 2.2 km for males (max = 22 km) and 3.2 km for females (max = 32 km). Approximately 70% of birds that nested in consecutive years on the study area had a different mate in the second year. Mate fidelity was related to box fidelity but not to prior nesting success or years of nesting experience. Mate changes occurred 32% of the time when the previous mate was known to be alive and nesting in the area. Kestrels that switched mates and boxes did not improve or decrease their subsequent nesting success. Kestrels usually switched to mates with less experience and lower lifetime productivity than their previous mates. The costs of switching boxes and mates were low, and there were no obvious benefits to fidelity. The cost of "waiting" for a previous mate that might have died could be high in species with high annual mortality.

Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

Science.gov (United States)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar

2012-01-01

A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

Fidel Castro: the word that unites

Directory of Open Access Journals (Sweden)

Grisel Veloz-Fernández

2016-09-01

Full Text Available The present work carries out a boarding of Fidel Castro's political speech, leaving of an understanding of its renovating character inside the most complex historical segment in the revolutionary process that is the period 1959 - 1961. It is during the same one that in Cuba a consent arms to the socialism. That achievement relapses fundamentally in Fidel's leadership and its capacity to negotiate the nucleus of ideas and values of that process through the political speech. Analyzed around 72 documents a characterization of that speech it has been achieved in benefit of their current utility. Presently work takes as starting point the conditions that existed in our country to the revolutionary victory that were the material base of an authentic revolution of the political speech in Cuba and the world like interaction form and transmission of political ideas as regards talkative resources.

Thermal Signature Measurements for Ammonium Nitrate/Fuel Mixtures by Laser Heating.

Science.gov (United States)

Nazarian, Ashot; Presser, Cary

2016-01-10

Measurements were carried out to obtain thermal signatures of several ammonium nitrate/fuel (ANF) mixtures, using a laser-heating technique referred to as the laser-driven thermal reactor (LDTR). The mixtures were ammonium nitrate (AN)/kerosene, AN/ethylene glycol, AN/paraffin wax, AN/petroleum jelly, AN/confectioner's sugar, AN/cellulose (tissue paper), nitromethane/cellulose, nitrobenzene/cellulose, AN/cellulose/nitromethane, AN/cellulose/nitrobenzene. These mixtures were also compared with AN/nitromethane and AN/diesel fuel oil, obtained from an earlier investigation. Thermograms for the mixtures, as well as individual constituents, were compared to better understand how the sample thermal signature changes with mixture composition. This is the first step in development of a thermal-signature database, to be used along with other signature databases, to improve identification of energetic substances of unknown composition. The results indicated that each individual thermal signature was associated unambiguously with a particular mixture composition. The signature features of a particular mixture were shaped by the individual constituent signatures. It was also uncovered that the baseline signature was modified after an experiment due to coating of unreacted residue on the substrate surface and a change in the reactor sphere oxide layer. Thus, care was required to pre-oxidize the sphere prior to an experiment. A minimum sample mass (which was dependent on composition) was required to detect the signature characteristics. Increased laser power served to magnify signal strength while preserving the signature features. For the mixtures examined, the thermal response of each ANF mixture was found to be different, which was based on the mixture composition and the thermal behavior of each mixture constituent.

Thermal System Modeling for Lunar and Martian Surface Regenerative Fuel Cell Systems

Science.gov (United States)

Gilligan, Ryan Patrick; Smith, Phillip James; Jakupca, Ian Joseph; Bennett, William Raymond; Guzik, Monica Christine; Fincannon, Homer J.

2017-01-01

The Advanced Exploration Systems (AES) Advanced Modular Power Systems (AMPS) Project is investigating different power systems for various lunar and Martian mission concepts. The AMPS Fuel Cell (FC) team has created two system-level models to evaluate the performance of regenerative fuel cell (RFC) systems employing different fuel cell chemistries. Proton Exchange Membrane fuel cells PEMFCs contain a polymer electrolyte membrane that separates the hydrogen and oxygen cavities and conducts hydrogen cations (protons) across the cell. Solid Oxide fuel cells (SOFCs) operate at high temperatures, using a zirconia-based solid ceramic electrolyte to conduct oxygen anions across the cell. The purpose of the modeling effort is to down select one fuel cell chemistry for a more detailed design effort. Figures of merit include the system mass, volume, round trip efficiency, and electrolyzer charge power required. PEMFCs operate at around 60 degrees Celsius versus SOFCs which operate at temperatures greater than 700 degrees Celsius. Due to the drastically different operating temperatures of the two chemistries the thermal control systems (TCS) differ. The PEM TCS is less complex and is characterized by a single pump cooling loop that uses deionized water coolant and rejects heat generated by the system to the environment via a radiator. The solid oxide TCS has its own unique challenges including the requirement to reject high quality heat and to condense the steam produced in the reaction. This paper discusses the modeling of thermal control systems for an extraterrestrial RFC that utilizes either a PEM or solid oxide fuel cell.

Thermal characterizations of the paraffin wax/low density polyethylene blends as a solid fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Soojong; Moon, Heejang; Kim, Jinkon, E-mail: jkkim@kau.ac.kr

2015-08-10

Highlights: • Regression rate of blends fuel is higher than polymer fuel. • LDPE is an effective mixing ingredient for the combustion efficiency. • Blends fuel is a uniform mixture with two degradation steps. • LDPE plays a positive role for the low sensitivity to the thermal deformation • Blends with low LDPE content can be an effective fuel for hybrid rocket application. - Abstract: Thermal characterizations of a novel solid fuel for hybrid rocket application, based on the paraffin wax blends with low density polyethylene (LDPE) concentration of 5% (SF-5) and 10% (SF-10) were conducted. Both the increased regression rate in comparison with the polymeric fuel, and the improved combustion efficiency in comparison with the pure paraffin fuel reveal that the blend fuels achieve higher combustion performance. The morphology of the shape stabilized paraffin wax/LDPE blends was characterized by the scanning electron microscopy (SEM). Although the SEM observation indicated the blends have uniform mixtures, they showed two degradation steps confirming the immiscibility of components in the crystalline phase from thermogravimetric analysis (TGA). The differential scanning calorimeter (DSC) results showed that the melting temperature of LDPE in the blends decreased with an increase of paraffin wax content. The decreasing total specific melting enthalpy of blended fuels with decreasing paraffin wax content is in fairly good agreement with the additive rule. In thermomechanical analysis (TMA), the linear coefficient of thermal expansion (LCTE) seems to decrease with an increase of LDPE loading, however, the loaded LDPE do merely affect the LCTE in case of the blends with low LDPE concentration. It was found that a blend of low concentration of LDPE with a relatively high concentration of paraffin wax can lead to a potential novel fuel for rocket application, a contrary case with respect to the field of phase change materials (PCM) where a blend of high concentration

Thermal characterizations of the paraffin wax/low density polyethylene blends as a solid fuel

International Nuclear Information System (INIS)

Kim, Soojong; Moon, Heejang; Kim, Jinkon

2015-01-01

Highlights: • Regression rate of blends fuel is higher than polymer fuel. • LDPE is an effective mixing ingredient for the combustion efficiency. • Blends fuel is a uniform mixture with two degradation steps. • LDPE plays a positive role for the low sensitivity to the thermal deformation • Blends with low LDPE content can be an effective fuel for hybrid rocket application. - Abstract: Thermal characterizations of a novel solid fuel for hybrid rocket application, based on the paraffin wax blends with low density polyethylene (LDPE) concentration of 5% (SF-5) and 10% (SF-10) were conducted. Both the increased regression rate in comparison with the polymeric fuel, and the improved combustion efficiency in comparison with the pure paraffin fuel reveal that the blend fuels achieve higher combustion performance. The morphology of the shape stabilized paraffin wax/LDPE blends was characterized by the scanning electron microscopy (SEM). Although the SEM observation indicated the blends have uniform mixtures, they showed two degradation steps confirming the immiscibility of components in the crystalline phase from thermogravimetric analysis (TGA). The differential scanning calorimeter (DSC) results showed that the melting temperature of LDPE in the blends decreased with an increase of paraffin wax content. The decreasing total specific melting enthalpy of blended fuels with decreasing paraffin wax content is in fairly good agreement with the additive rule. In thermomechanical analysis (TMA), the linear coefficient of thermal expansion (LCTE) seems to decrease with an increase of LDPE loading, however, the loaded LDPE do merely affect the LCTE in case of the blends with low LDPE concentration. It was found that a blend of low concentration of LDPE with a relatively high concentration of paraffin wax can lead to a potential novel fuel for rocket application, a contrary case with respect to the field of phase change materials (PCM) where a blend of high concentration

High thermal efficiency and low emission performance of a methanol reformed gas fueled engine for hybrid electric vehicles

Energy Technology Data Exchange (ETDEWEB)

Yamane, K.; Nakajima, Y.; Shudo, T.; Hiruma, M. [Musahi Inst. of Tech., Tokyo (Japan); Komatsu, H.; Takagi, Y. [Nissan Motor Co., Ltd., Yokosuka (Japan)

2000-07-01

An internal combustion engine (ICE) operation was carried out experimentally by using the mixture of air and fuel simulating the reformed gas as the fuel. It has been found that the engine can expectedly attain ultra-low emission and high thermal efficiency, namely 35% brake thermal efficiency in the basis of the low heat value of the theoretically reformed gas or 42% in the basis of the low heat value of methanol. By using the result for the estimation of the total thermal efficiency at the end of the motor output shaft of a hybrid electric vehicle, it has been found that the total thermal efficiency of the reformed gas engine system is 34% in case of a 120% energy increment and 33% in case of a 116% energy increment with a little higher NOx emission of 60 ppm while the counterpart of the fuel cell system is 34%. When the emission level for EZEV is required, the total thermal efficiency falls to 32% in case of a 120% energy increment and 31% in case of a 116% energy increment. From the points of the reliability proved by the long history, higher specific power and low cost, the internal combustion engine system with the thermal efficiency almost equal to that of the fuel cell (FC) system is further more practical when methanol is used as the fuel. (orig.)

Thermal and in-pile densification of MOX fuels: Some recent results

International Nuclear Information System (INIS)

Caillot, L.; Malgouyres, P.P.; Souchon, F.; Gotta, M.J.; Warin, D.; Chotard, A.; Couty, J.C.

1997-01-01

In-pile densification of PWR fuels is one of the main phenomena which determine the evolution of the pellet-clad gap during the first stage of the irradiation, and thus has consequences onto the thermo-mechanical behaviours of fuel rods. It can be predicted using the results of resintering tests and appropriate correlations. In this context, CEA, FRAMATOME and EDF have undertaken a joint research programme aiming to characterize the densification of MOX fuels. Different fuels were prepared by the MIMAS process using different UO 2 powders as matrix. After a detailed characterization, fuel pellets were submitted to isothermal resintering tests and analytical irradiations. Correlations between in-pile and thermal densification were established. This paper presents the results obtained with two types of MOX fuel: one fabricated wit the AUC UO 2 powder (ammonium uranyl carbonate conversion process) and another one fabricated with the SFEROX powder (peroxide conversion process). 8 refs, 8 figs

Conjugate heat transfer simulations of advanced research reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Piro, M.H.A., E-mail: pirom@aecl.ca; Leitch, B.W.

2014-07-01

Highlights: • Temperature predictions are enhanced by coupling heat transfer in solid and fluid zones. • Seven different cases are considered to observe trends in predicted temperature and pressure. • The seven cases consider high/medium/low power, flow, burnup, fuel material and geometry. • Simulations provide temperature predictions for performance/safety. Boiling is unlikely. • Simulations demonstrate that a candidate geometry can enhance performance/safety. - Abstract: The current work presents numerical simulations of coupled fluid flow and heat transfer of advanced U–Mo/Al and U–Mo/Mg research reactor fuels in support of performance and safety analyses. The objective of this study is to enhance predictions of the flow regime and fuel temperatures through high fidelity simulations that better capture various heat transfer pathways and with a more realistic geometric representation of the fuel assembly in comparison to previous efforts. Specifically, thermal conduction, convection and radiation mechanisms are conjugated between the solid and fluid regions. Also, a complete fuel element assembly is represented in three dimensional space, permitting fluid flow and heat transfer to be simulated across the entire domain. Seven case studies are examined that vary the coolant inlet conditions, specific power, and burnup to investigate the predicted changes in the pressure drop in the coolant and the fuel, clad and coolant temperatures. In addition, an alternate fuel geometry is considered with helical fins (replacing straight fins in the existing design) to investigate the relative changes in predicted fluid and solid temperatures. Numerical simulations predict that the clad temperature is sensitive to changes in the thermal boundary layer in the coolant, particularly in simultaneously developing flow regions, while the temperature in the fuel is anticipated to be unaffected. Finally, heat transfer between fluid and solid regions is enhanced with

IFPE/IFA-533, Fuel Thermal Behaviour at High Burnup, Halden Reactor

International Nuclear Information System (INIS)

Gyori, Cs.; Turnbull, J.A.

1997-01-01

Description: After twelve years irradiation in the Halden Boiling Water Reactor two fuel rods (Rod 807 and Rod 808) were re-instrumented with fuel centre thermocouples and reloaded into the reactor in order to investigate the fuel thermal behaviour at high burnup. The fuel rods were pre-irradiated with four other rods in the upper cluster of IFA-409 (IFA=Instrumented Fuel Assembly) from May 1973 to June 1985. After base irradiation the four neighbouring rods were re-instrumented with pressure transducers and ramp tested in IFA-535.5 and IFA-535.6 providing useful data about fission gas release (FGR) presented in the Fuel Performance Database as well (Ref. 1). The two rods re-instrumented with fuel centre thermocouples have been irradiated as IFA-533.2 from April 1992. As the irradiation history of IFA-533.2 in the first months was very similar to the history of the ramp tests, the fuel temperature and FGR data measured in the different IFAs can complement each other, although the fuel-cladding gap sizes were slightly different and due to re-instrumentation the internal gas conditions were also dissimilar

A comparative analysis of the effect of gaseous fission products release on the thermal behaviour of oxide fuel rods

International Nuclear Information System (INIS)

Totev, T.L.; Kolev, I.G.

1992-01-01

Four different models of gaseous fission product release are compared in order to assess the relative effect of thermal characteristics of the fuel rods. The results show that the use of Weisman and EPRI models at a high burnup (over 50000 MW.d/tU) leads to almost the same figures of maximum fuel temperature and gas gap thermal conductivity. The use of Beyer-Hann (Betelle) and Pazdera-Valach (Rzez) models leads to under prediction of the fuel element thermal characteristics. A conclusion has been made that the Weisman model is the most suitable for the WWER-type fuel elements behaviour prediction. 10 refs., 7 figs

C.E.C. - cod for calculus of the evolution fuel for thermal reactors

International Nuclear Information System (INIS)

Biciolla, L.; Marcu, G.; Mociornita, G.

1975-01-01

The study of ''burnup'' into thermal reactor involves two main aspects: the economic one and another regarding the reactor operation, its stability and control. In the CEC-code written in FORTRAN IV language was analysed the change of the isotopic composition of nuclear fuel from thermal reactor during its operation

Thermal stress analysis of the fuel storage facility

International Nuclear Information System (INIS)

Chen, W.W.

1991-12-01

This paper presents the results of a nonlinear finite-element analysis to determine the structural integrity of the walls of the nuclear fuel storage room in the Radio Isotope Power System Facility of the Fuels and Materials Examination Facility (FMEF) Project. The analysis was performed to assess the effects of thermal loading on the walls that would result from a loss-of-cooling accident. The results obtained from using the same three-dimensional finite-element model with different types of elements, the eight-node brick element and the nonlinear concrete element, and the calculated results using the analytical solutions, are compared. The concrete responses in terms of octahedral normal and shearing stresses are described. The crack and crush states of the concrete were determined on the basis of multiaxial failure criteria

Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

2016-12-15

The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

Dynamic Thermal Model And Control Of A Pem Fuel Cell System

DEFF Research Database (Denmark)

Liso, Vincenzo; Nielsen, Mads Pagh

2013-01-01

the fuel cell system. A PID temperature control is implemented to study the effect of stack temperature on settling times of other variables such as stack voltage, air flow rate, oxygen excess ratio and net power of the stack. The model allows an assessment of the effect of operating parameters (stack...... power output, cooling water flow rate, air flow rate, and environmental temperature) and parameter interactions on the system thermal performance. The model represents a useful tool to determine the operating temperatures of the various components of the thermal system, and thus to fully assess......A lumped parameter dynamic model is developed for predicting the stack performance, temperatures of the exit reactant gases and coolant liquid outlet in a proton-exchange membrane fuel cell (PEMFC) system. The air compressor, humidifier and cooling heat exchanger models are integrated to study...

Theoretical basis for a transient thermal elastic-plastic stress analysis of nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hsu, T.R.; Bertels, A.W.M.; Banerjee, S.; Harrison, W.C.

1976-07-01

This report presents the theoretical basis for a transient thermal elastic-plastic stress analysis of a nuclear reactor fuel element subject to severe transient thermo-mechanical loading. A finite element formulation is used for both the non-linear stress analysis and thermal analysis. These two major components are linked together to form an integrated program capable of predicting fuel element transient behaviour in two dimensions. Specific case studies are presented to illustrate capabilities of the analysis. (author)

Fuels and chemicals from biomass using solar thermal energy

Science.gov (United States)

Giori, G.; Leitheiser, R.; Wayman, M.

1981-01-01

The significant nearer term opportunities for the application of solar thermal energy to the manufacture of fuels and chemicals from biomass are summarized, with some comments on resource availability, market potential and economics. Consideration is given to the production of furfural from agricultural residues, and the role of furfural and its derivatives as a replacement for petrochemicals in the plastics industry.

Fidelity deviation in quantum teleportation

Science.gov (United States)

Bang, Jeongho; Ryu, Junghee; Kaszlikowski, Dagomir

2018-04-01

We analyze the performance of quantum teleportation in terms of average fidelity and fidelity deviation. The average fidelity is defined as the average value of the fidelities over all possible input states and the fidelity deviation is their standard deviation, which is referred to as a concept of fluctuation or universality. In the analysis, we find the condition to optimize both measures under a noisy quantum channel—we here consider the so-called Werner channel. To characterize our results, we introduce a 2D space defined by the aforementioned measures, in which the performance of the teleportation is represented as a point with the channel noise parameter. Through further analysis, we specify some regions drawn for different channel conditions, establishing the connection to the dissimilar contributions of the entanglement to the teleportation and the Bell inequality violation.

Method of manufacturing nuclear fuel pellet

International Nuclear Information System (INIS)

Oguma, Masaomi; Masuda, Hiroshi; Hirai, Mutsumi; Tanabe, Isami; Yuda, Ryoichi.

1989-01-01

In a method of manufacturing nuclear fuel pellets by compression molding an oxide powder of nuclear fuel material followed by sintering, a metal nuclear material is mixed with an oxide powder of the nuclear fuel material. As the metal nuclear fuel material, whisker or wire-like fine wire or granules of metal uranium can be used effectively. As a result, a fuel pellet in which the metal nuclear fuel is disposed in a network-like manner can be obtained. The pellet shows a great effect of preventing thermal stress destruction of pellets upon increase of fuel rod power as compared with conventional pellets. Further, the metal nuclear fuel material acts as an oxygen getter to suppress the increase of O/M ratio of the pellets. Further, it is possible to reduce the swelling of pellet at high burn-up degree. (T.M.)

KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

International Nuclear Information System (INIS)

Shamasundar, B.I.; Fehrenbach, M.E.

1981-05-01

The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

Thermal Mode of Tanks for Storage Fuel of Thermal Power Plants and Boiler with the Influence of Engineering Facilities in the Area of their Placement

Science.gov (United States)

Polovnikov, V. Yu.; Makhsutbek, F. T.; Ozhikenova, Zh. F.

2016-02-01

This paper describes the numerical modeling of heat transfer in the area placing of the tank for storage fuel of thermal power plant and boiler with the influence of engineering construction. We have established that the presence of engineering structures in the area of placing of the tank for storage fuel of thermal power plant and boiler have little effect on the change of heat loss.

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Energy Technology Data Exchange (ETDEWEB)

El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

2017-11-15

In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

THE CONCEPT OF FIDELITY IN COMICS TRANSLATION

Directory of Open Access Journals (Sweden)

Erico Assis

2016-11-01

Full Text Available The long-discussed – and frequently dismissed – concept of translation faithfulness or translation fidelity, though usually applied to literary texts, has its fair share of applications when considered for comics translation. In literary translation, non-linguistic portions such as illustrations are often considered addenda or “paratexts” relative to the main, linguistic text. Comics, by its turn, present a certain set of features which single them out as a form that demands a new concept of “text” and, therefore, of translation fidelity. The comic-reading process, as pertaining to cognitive apprehension, implies interpretative accords that differ from the ones in purely linguistic texts: each and every element of the comics page – non-linguistic (mainly imagetic signs, linguistic signs, panel borders, typography and such – are intertwined and should be perceived in regards to its spatial and topological relations. This approach to understanding comics is based on Groensteen (1999 and his concepts of arthrology, spatio-topia, page layout, breakdown and braiding. As for translation fidelity, we rely on authors such as Berman (1984, Guidere (2010 and Aubert (1993. On comics translation, Zanettin (2008, Rota (2008 and Yuste Frías (2010, 2011 are of particular interest. Based on various concepts of fidelity – supported by samples of translated comics with varied degrees of fidelity to the source text – we discuss the different grounds of source-text fidelity, target-reader fidelity and source-author fidelity in the following instances: linguistic sign fidelity, imagetic sign fidelity, spatio-topia fidelity, typographic fidelity and format fidelity.

High fidelity nuclear energy system optimization towards an environmentally benign, sustainable, and secure energy source

International Nuclear Information System (INIS)

Tsvetkov, Pavel Valeryevich; Rodriguez, Salvador B.; Ames, David E. II; Rochau, Gary Eugene

2010-01-01

A new high-fidelity integrated system method and analysis approach was developed and implemented for consistent and comprehensive evaluations of advanced fuel cycles leading to minimized Transuranic (TRU) inventories. The method has been implemented in a developed code system integrating capabilities of Monte Carlo N - Particle Extended (MCNPX) for high-fidelity fuel cycle component simulations. In this report, a Nuclear Energy System (NES) configuration was developed to take advantage of used fuel recycling and transmutation capabilities in waste management scenarios leading to minimized TRU waste inventories, long-term activities, and radiotoxicities. The reactor systems and fuel cycle components that make up the NES were selected for their ability to perform in tandem to produce clean, safe, and dependable energy in an environmentally conscious manner. The diversity in performance and spectral characteristics were used to enhance TRU waste elimination while efficiently utilizing uranium resources and providing an abundant energy source. A computational modeling approach was developed for integrating the individual models of the NES. A general approach was utilized allowing for the Integrated System Model (ISM) to be modified in order to provide simulation for other systems with similar attributes. By utilizing this approach, the ISM is capable of performing system evaluations under many different design parameter options. Additionally, the predictive capabilities of the ISM and its computational time efficiency allow for system sensitivity/uncertainty analysis and the implementation of optimization techniques.

High fidelity nuclear energy system optimization towards an environmentally benign, sustainable, and secure energy source.

Energy Technology Data Exchange (ETDEWEB)

Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Rodriguez, Salvador B.; Ames, David E., II (Texas A& M University, College Station, TX); Rochau, Gary Eugene

2010-10-01

A new high-fidelity integrated system method and analysis approach was developed and implemented for consistent and comprehensive evaluations of advanced fuel cycles leading to minimized Transuranic (TRU) inventories. The method has been implemented in a developed code system integrating capabilities of Monte Carlo N - Particle Extended (MCNPX) for high-fidelity fuel cycle component simulations. In this report, a Nuclear Energy System (NES) configuration was developed to take advantage of used fuel recycling and transmutation capabilities in waste management scenarios leading to minimized TRU waste inventories, long-term activities, and radiotoxicities. The reactor systems and fuel cycle components that make up the NES were selected for their ability to perform in tandem to produce clean, safe, and dependable energy in an environmentally conscious manner. The diversity in performance and spectral characteristics were used to enhance TRU waste elimination while efficiently utilizing uranium resources and providing an abundant energy source. A computational modeling approach was developed for integrating the individual models of the NES. A general approach was utilized allowing for the Integrated System Model (ISM) to be modified in order to provide simulation for other systems with similar attributes. By utilizing this approach, the ISM is capable of performing system evaluations under many different design parameter options. Additionally, the predictive capabilities of the ISM and its computational time efficiency allow for system sensitivity/uncertainty analysis and the implementation of optimization techniques.

Advanced thermally stable jet fuels. Technical progress report, April 1993--June 1993

Energy Technology Data Exchange (ETDEWEB)

Schobert, H.H.; Eser, S.; Song, C. [and others

1993-10-01

The Penn State program in advanced thermally stable coal-based jet fuels has five broad objectives: (1) development of mechanisms of degradation and solids formation; (2) quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) elucidation of the role of additives in retarding the formation of carbonaceous solids; and (5) assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Some of our accomplishments and findings are: The product distribution and reaction mechanisms for pyrolysis of alkylcyclohexanes at 450{degree}C have been investigated in detail. In this report we present results of pyrolysis of cyclohexane and a variety of alkylcyclohexanes in nitrogen atmospheres, along with pseudo-first order rate constants, and possible reaction mechanisms for the origin of major pyrolysis products are presented. Addition of PX-21 activated carbon effectively stops the formation of carbonaceous solids on reactor walls during thermal stressing of JPTS. A review of physical and chemical interactions in supercritical fluids has been completed. Work has begun on thermal stability studies of a second generation of fuel additives, 1,2,3,4-tetrahydro-l-naphthol, 9,10-phenanthrenediol, phthalan, and 1,2-benzenedimethanol, and with careful selection of the feedstock, it is possible to achieve 85--95% conversion of coal to liquids, with 40--50% of the dichloromethane-soluble products being naphthalenes. (Further hydrogenation of the naphthalenes should produce the desired highly stable decalins.)

Thermal balance of a four stroke SI engine operating on hydrogen as a supplementary fuel

International Nuclear Information System (INIS)

Yueksel, F.; Ceviz, M.A.

2003-01-01

This paper investigates the effects of adding constant quantity hydrogen to gasoline-air mixture on SI engine thermal balance and performance. A four stroke, four-cylinder SI engine was used for conducting this study. Thermal balance tests were conducted for engine thermal efficiency, heat loss through the exhaust gases, heat loss to the cooling water and unaccounted losses (i.e. heat lost by lubricating oil, radiation), while performance tests were in respect to the brake power, specific fuel consumption and air ratio. Hydrogen supplementations were used with three different and fixed mass flow rates; 0.129, 0.168 and 0.208 kg h -1 at near three-fourth throttle opening position and variable engine speed ranging from 1000 to 4500 rpm. The results showed that supplementation of hydrogen to gasoline decreases the heat loss to cooling water and unaccounted losses, and the heat loss through the exhaust gas is nearly the same with pure gasoline experiments. Additionally, specific fuel consumption decreases, while the engine thermal efficiency and the air ratio increase. Engine performance parameters such as thermal efficiency and specific fuel consumption improved the level of the ratio of hydrogen mass flow rate to that of gasoline up to 5%

Non-fertile fuels for burning weapons plutonium in thermal fission reactors

International Nuclear Information System (INIS)

Lombardi, C.; Mazzola, A.; Vettraino, F.

1996-01-01

In the last few years, the excess plutonium disposition has become ever more a topical and critical issue. As a matter of fact, more than 200 MT of plutonium coming from spent fuel reprocessing have been already stockpiled and over the next decade, under the already ratified agreements, another about 200 MT of weapon-grade plutonium are expected to be available from nuclear weapons dismantlement. On this basis, an ever growing plutonium production is no longer the goal and the already stored quantities should be burnt in power reactors by taking care that no new plutonium is generated under irradiation. This new outlook in considering plutonium has led many designers to reassess the Fast Breeder Reactors (FBR) role and shifting from breeder to burner machines perspective. Several solutions for burning plutonium have been so far proposed and discussed from the safeguards, proliferation resistance, environmental safety, technological background, economy and time schedule standpoint. A proposal for plutonium burning in commercial Pressurized Water Reactors (PWR) by using a non-fertile oxide-type fuel consisting of PuO 2 diluted in an inert matrix is reported hereafter. This solution appears to receive an ever growing interest in the nuclear community. In order not to produce new plutonium during irradiation an innovative U-free fuel is being researched, based on an inert matrix which will consist in a mixed compound of inert oxides, such as ZrO 2 , Al2O 3 , MgO, CeO 2 where the plutonium oxide is dispersed in. The matrix will fulfill the following requirements: good chemical compatibility, acceptable thermal conductivity, good nuclear properties, good stability under irradiation, good dissolution resistance. The plutonium relative content will be comparable to that used in MOX fuel. The fuel is expected to be characterized by a high chemical stability (rock-like fuel), so that after discharge from reactor and adequate cooling time, it can be considered a High Level

Hydrogen Fuel as Ecological Contribution to Operation of the Existing Coal-Fired Thermal Power Plants

International Nuclear Information System (INIS)

Cosic, D.

2009-01-01

The analysis is carried out of the application of a new hydrogen based alternative fuel as ecological contribution of the coal thermal power plants operation. Given the fact that coal thermal power plants are seen as the largest producers, not only of CO 2 , but of all others harmful gases, the idea is initiated to use the new alternative fuel as an additive to the coal which would result in much better performance of the coal power plants from an ecological point of view. It is possible to use such a fuel in relation of 10-30% of former coal use. The positive influence of such an application is much bigger than relative used quantity. This lecture has a goal to incite potential investors to create conditions for industrial testing of the new fuel. It will be very interesting to animate investors for large-scale production of the new fuel, too.(author).

Study on thermal conductivity of HTR spherical fuel element matrix graphite

International Nuclear Information System (INIS)

Zhang Kaihong; Liu Xiaoxue; Zhao Hongsheng; Li Ziqiang; Tang Chunhe

2014-01-01

Taking the spherical fuel element matrix graphite ball samples as an example, this paper introduced the principle and method of laser thermal conductivity meter, as well as the specific heat capacity, and analyzed the effects of different test methods and sampling methods on the thermal conductivities at 1000 ℃ of graphite material. The experimental results show that the thermal conductivities of graphite materials tested by synchronous thermal analyzer combining with laser thermal conductivity meter were different from that directly by laser thermal conductivity meter, the former was more reliable and accurate than the later; When sampling from different positions, central samples had higher thermal conductivities than edging samples, which was related to the material density and porosity at the different locations; the thermal conductivities had obvious distinction between samples from different directions, which was because the layer structure of polycrystalline graphite preferred orientation under pressure, generally speaking, the thermal conductivities perpendicular to the molding direction were higher than that parallel to the molding direction. Besides this, the test results show that the thermal conductivities of all the graphite material samples were greater than 30 W/(m (K), achieving the thermal performance index of high temperature gas cooled reactor. (authors)

Sol-gel process for thermal reactor fuel fabrication

International Nuclear Information System (INIS)

Mukerjee, S.K.

2008-01-01

Full text: Sol-gel processes have revolutionized conventional ceramic technology by providing extremely fine and uniform powders for the fabrication of ceramics. The use of this technology for nuclear fuel fabrication has also been explored in many countries. Unlike the conventional sol-gel process, sol-gel process for nuclear fuels tries to eliminate the preparation of powders in view of the toxic nature of the powders particularly those of plutonium and 233 U. The elimination of powder handling thus makes this process more readily amenable for use in glove boxes or for remote handling. In this process, the first step is the preparation of microspheres of the fuel material from a solution which is then followed by vibro-compaction of these microspheres of different sizes to obtain the required smear density of fuel inside a pin. The maximum achievable packing density of 92 % makes it suitable for fast reactors only. With a view to extend the applicability of sol-gel process for thermal reactor fuel fabrication the concept of converting the gel microspheres derived from sol-gel process, to the pellets, has been under investigation for several years. The unique feature of this process is that it combines the advantages of sol-gel process for the preparation of fuel oxide gel microspheres of reproducible quality with proven irradiation behavior of the pellet fuel. One of the important pre-requisite for the success of this process is the preparation of soft oxide gel microspheres suitable for conversion to dense pellets free from berry structure. Studies on the internal gelation process, one of the many variants of sol-gel process, for obtaining soft oxide gel microspheres suitable for gel pelletisation is now under investigation at BARC. Some of the recent findings related to Sol-Gel Microsphere Pelletisation (SGMP) in urania-plutonia and thoria-urania systems will be presented

PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

International Nuclear Information System (INIS)

Ganzmann, I.; Hille, D.; Staude, U.

2009-01-01

The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

Effects of water-emulsified fuel on a diesel engine generator's thermal efficiency and exhaust.

Science.gov (United States)

Syu, Jin-Yuan; Chang, Yuan-Yi; Tseng, Chao-Heng; Yan, Yeou-Lih; Chang, Yu-Min; Chen, Chih-Chieh; Lin, Wen-Yinn

2014-08-01

Water-emulsified diesel has proven itself as a technically sufficient improvement fuel to improve diesel engine fuel combustion emissions and engine performance. However, it has seldom been used in light-duty diesel engines. Therefore, this paper focuses on an investigation into the thermal efficiency and pollution emission analysis of a light-duty diesel engine generator fueled with different water content emulsified diesel fuels (WD, including WD-0, WD-5, WD-10, and WD-15). In this study, nitric oxide, carbon monoxide, hydrocarbons, and carbon dioxide were analyzed by a vehicle emission gas analyzer and the particle size and number concentration were measured by an electrical low-pressure impactor. In addition, engine loading and fuel consumption were also measured to calculate the thermal efficiency. Measurement results suggested that water-emulsified diesel was useful to improve the thermal efficiency and the exhaust emission of a diesel engine. Obviously, the thermal efficiency was increased about 1.2 to 19.9%. In addition, water-emulsified diesel leads to a significant reduction of nitric oxide emission (less by about 18.3 to 45.4%). However the particle number concentration emission might be increased if the loading of the generator becomes lower than or equal to 1800 W. In addition, exhaust particle size distributions were shifted toward larger particles at high loading. The consequence of this research proposed that the water-emulsified diesel was useful to improve the engine performance and some of exhaust emissions, especially the NO emission reduction. Implications: The accumulated test results provide a good basis to resolve the corresponding pollutants emitted from a light-duty diesel engine generator. By measuring and analyzing transforms of exhaust pollutant from this engine generator, the effects of water-emulsified diesel fuel and loading on emission characteristics might be more clear. Understanding reduction of pollutant emissions during the use

Optimal initial fuel distribution in a thermal reactor for maximum energy production

International Nuclear Information System (INIS)

Moran-Lopez, J.M.

1983-01-01

Using the fuel burnup as objective function, it is desired to determine the initial distribution of the fuel in a reactor in order to obtain the maximum energy possible, for which, without changing a fixed initial fuel mass, the results for different initial fuel and control poison configurations are analyzed and the corresponding running times compared. One-dimensional, two energy-group theory is applied to a reflected cylindrical reactor using U-235 as fuel and light water as moderator and reflector. Fissions in both fast and thermal groups are considered. The reactor is divided into several annular regions, and the constant flux approximation in each depletion step is then used to solve the fuel and fission-product poisons differential equations in each region. The computer code OPTIME was developed to determine the time variation of core properties during the fuel cycle. At each depletion step, OPTIME calls ODMUG, [12] a criticality search program, from which the spatially-averaged neutron fluxes and control poison cross sections are obtained

Simulation of Thermal, Neutronic and Radiation Characteristics in Spent Nuclear Fuel and Radwaste Facilities

International Nuclear Information System (INIS)

Poskas, P.; Bartkus, G.

1999-01-01

The overview of the activities in the Division of Thermo hydro-mechanics related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Also some new data about radiation characteristics of the RBMK-1500 spent nuclear fuel are presented. (author)

Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

International Nuclear Information System (INIS)

Gwaltney, R.C.; McAfee, W.J.

1977-01-01

The equivalent solid plate method, in conjunction with two-dimensional plane stress and plane strain analyses, was used in assessing the thermal stress behavior of HTGR fuel and control rod fuel blocks. For the control rod fuel blocks, particular attention was given to ascertaining the effects of the reserve shutdown hole and the control rod channel holes. The assumed safety factor of 2 on the failure criteria was considered adequate to account for neglecting the axial temperature gradient in the plane analyses of the ends of the blocks. The analyses indicated that the maximum calculated tensile stress values were smaller than the criteria values except for the plane strain analysis of the control rod fuel block end surfaces and the axisymmetric analysis of the fuel block as a circular cylinder. However, most of the maximum calculated strain values were greater than the criteria values

Verification of the thermal module in the ELESIM code and the associated uncertainty analysis

International Nuclear Information System (INIS)

Arimescu, V.I.; Williams, A.F.; Klein, M.E.; Richmond, W.R.; Couture, M.

1997-09-01

Temperature is a critical parameter in fuel modelling because most of the physical processes that occur in fuel elements during irradiation are thermally activated. The focus of this paper is the temperature distribution calculation used in the computer code ELESIM, developed at AECL to model the steady-state behaviour of CANDU fuel. A validation procedure for fuel codes is described and applied to ELESIM's thermal calculation.The effects of uncertainties in model parameters, like Uranium Dioxide thermal conductivity, and input variables, such as fuel element linear power, are accounted for through an uncertainty analysis using Response Surface and Monte Carlo techniques

Comparison of thermal compatibility between atomized and comminuted U3Si dispersion fuels

International Nuclear Information System (INIS)

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu; Kuk, II-Hyun

1997-01-01

Thermal compatibility of atomized U 3 Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500 degrees C, and compared with that of comminuted U 3 Si. Atomized U 3 Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U 3 Si and Al occurred along the grain boundaries and deformation bands in U 3 Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, and reduced the volume expansion

Methods and systems to thermally protect fuel nozzles in combustion systems

Science.gov (United States)

Helmick, David Andrew; Johnson, Thomas Edward; York, William David; Lacy, Benjamin Paul

2013-12-17

A method of assembling a gas turbine engine is provided. The method includes coupling a combustor in flow communication with a compressor such that the combustor receives at least some of the air discharged by the compressor. A fuel nozzle assembly is coupled to the combustor and includes at least one fuel nozzle that includes a plurality of interior surfaces, wherein a thermal barrier coating is applied across at least one of the plurality of interior surfaces to facilitate shielding the interior surfaces from combustion gases.

Nuclear power plant training simulator fidelity assessment

International Nuclear Information System (INIS)

Carter, R.J.; Laughery, K.R.

1985-01-01

The fidelity assessment portion of a methodology for evaluating nuclear power plant simulation facilities in regard to their appropriateness for conducting the Nuclear Regulatory Commission's operating test was described. The need for fidelity assessment, data sources, and fidelity data to be collected are addressed. Fidelity data recording, collection, and analysis are discussed. The processes for drawing conclusions from the fidelity assessment and evaluating the adequacy of the simulator control-room layout were presented. 3 refs

Determination of thermal reactivity coefficients for the first fuel loading of MO34

International Nuclear Information System (INIS)

Lueley, J.; Vrban, B.; Farkas, G.; Hascik, J.; Hinca, R.; Petriska, M.; Slugen, V.

2012-01-01

The article introduces determination of thermal reactivity coefficients, especially summarized (isothermal) and moderator (density) reactivity coefficients between 200 grad C and 260 grad C with 2 grad C step, - in compliance with the assignment - for the first fuel loading into the RC of NP Mochovce units using 2 nd generation fuel during the start-up using calculation code MCNP5 1.60. (authors)

Physiological Based Simulator Fidelity Design Guidance

Science.gov (United States)

Schnell, Thomas; Hamel, Nancy; Postnikov, Alex; Hoke, Jaclyn; McLean, Angus L. M. Thom, III

2012-01-01

The evolution of the role of flight simulation has reinforced assumptions in aviation that the degree of realism in a simulation system directly correlates to the training benefit, i.e., more fidelity is always better. The construct of fidelity has several dimensions, including physical fidelity, functional fidelity, and cognitive fidelity. Interaction of different fidelity dimensions has an impact on trainee immersion, presence, and transfer of training. This paper discusses research results of a recent study that investigated if physiological-based methods could be used to determine the required level of simulator fidelity. Pilots performed a relatively complex flight task consisting of mission task elements of various levels of difficulty in a fixed base flight simulator and a real fighter jet trainer aircraft. Flight runs were performed using one forward visual channel of 40 deg. field of view for the lowest level of fidelity, 120 deg. field of view for the middle level of fidelity, and unrestricted field of view and full dynamic acceleration in the real airplane. Neuro-cognitive and physiological measures were collected under these conditions using the Cognitive Avionics Tool Set (CATS) and nonlinear closed form models for workload prediction were generated based on these data for the various mission task elements. One finding of the work described herein is that simple heart rate is a relatively good predictor of cognitive workload, even for short tasks with dynamic changes in cognitive loading. Additionally, we found that models that used a wide range of physiological and neuro-cognitive measures can further boost the accuracy of the workload prediction.

Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

Energy Technology Data Exchange (ETDEWEB)

Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

2017-07-01

To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

International Nuclear Information System (INIS)

Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

2017-01-01

To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

Features of the Numerical Solution of Thermal Destruction Fuel Pins Problems in the Fast Reactor

Science.gov (United States)

Usov, E. V.; Butov, A. A.; Klimonov, I. A.; Chuhno, V. I.; Nikolaenko, A. V.; Zhdanov, V. S.; Pribaturin, N. A.; Strizhov, V. F.

2017-11-01

In this paper the description of the basic equations which can be used for calculation of melting of fuel and cladding of the fast reactor, moving of the melt on a fuel pin surface and its solidification is presented. The special attention is given speed of calculation algorithms and fidelity of the phenomena which are observed at a stage of severe accidents in fast reactors. For check of working capacity of initial models, numerical calculations of Stefan-type problems on front movement of melting/solidification in cylindrical geometry are presented. Comparison with the solutions received by known analytical methods is executed. For validation of the numerical realization of calculation algorithms the analysis is carried out and experiments in which melting of the model fuel pins of fast reactors was studied are chosen. On the basis of the chosen experiments calculation schemes taking into account initial and boundary conditions are prepared and modeling is performed. Modeling results are shown in the present paper. Estimation of calculation error of the basic physical parameters is done by results of the modeling and conclusions are drawn on a correctness of algorithms operation.

Thermal Properties of Green Fuel Briquettes from Residue Corncobs Materials Mixed Macadamia Shell Charcoal Powder

Science.gov (United States)

Teeta, Suminya; Nachaisin, Mali; Wanish, Suchana

2017-09-01

The objective of this research was to produce green fuel briquettes from corncobs by adding macadamia shell charcoal powder. The study was sectioned into 3 parts: 1) Quality improvement of green fuel briquettes by adding macadamia; 2) Fuel property analysis based on ASTM standards and thermal fuel efficiency; and 3) Economics appropriateness in producing green fuel briquettes. This research produced green fuel briquettes using the ratio of corncobs weight and macadamia shell charcoal powder in 100:0 90:10 80:20 70:30 60:40 and 50:50 and pressing in the cold briquette machine. Fuel property analysis showed that green fuel briquettes at the ratio 50:50 produced maximum heating values at 21.06 Megajoule per kilogram and briquette density of 725.18 kilograms per cubic meter, but the percent of moisture content, volatile matter, ash, and fixed carbon were 10.09, 83.02, 2.17 and 4.72 respectively. The thermal efficiency of green fuel briquettes averaged 20.22%. Economics appropriateness was most effective where the ratio of corncobs weight to macadamia shell charcoal powder was at 50:50 which accounted for the cost per kilogram at 5.75 Baht. The net present value was at 1,791.25 Baht. Internal rate of return was at 8.62 and durations for a payback period of investment was at 1.9 years which was suitable for investment.

Fuel Design for Particle-Bed Reactors for Thermal Propulsion Applications

Science.gov (United States)

Husser, Dewayne L.; Evans, Robert S.; Jensen, Russell R.; Kerr, John M.

1994-07-01

The design of particle bed reactor (PBR) fuels is an iterative process involving close coordination of design and manufacturing operations. The process starts with the generation of an initial particle design, based on a knowledge of the system requirements and interfaces (such as, fissile loading requirements, coolant type, exit gas temperatures, operation time, number of cycles, contacting materials, etc.). The designer must consider materials property data, heat-transfer and thermal-hydraulic characteristics of the particle and particle bed, and available (or anticipated) manufacturing technology. The design process also uses parametric studies to identify the influences of composition, size, and coating thickness on fuel performance. This resulting design is then used to provide a target manufacturing specification against which initial manufacturing development can be assessed and which provides the framework for manufacturing and testing derived feedback that can be incorporated into the subsequent particle design modifications. In this paper, an example of this design process for a hypothetical particle using a (U,Zr)C kernel and a NbC outer coating designed for a thermal propulsion application is given.

Assessing fidelity of delivery of smoking cessation behavioural support in practice.

Science.gov (United States)

Lorencatto, Fabiana; West, Robert; Christopherson, Charlotte; Michie, Susan

2013-04-04

Effectiveness of evidence-based behaviour change interventions is likely to be undermined by failure to deliver interventions as planned. Behavioural support for smoking cessation can be a highly cost-effective, life-saving intervention. However, in practice, outcomes are highly variable. Part of this may be due to variability in fidelity of intervention implementation. To date, there have been no published studies on this. The present study aimed to: evaluate a method for assessing fidelity of behavioural support; assess fidelity of delivery in two English Stop-Smoking Services; and compare the extent of fidelity according to session types, duration, individual practitioners, and component behaviour change techniques (BCTs). Treatment manuals and transcripts of 34 audio-recorded behavioural support sessions were obtained from two Stop-Smoking Services and coded into component BCTs using a taxonomy of 43 BCTs. Inter-rater reliability was assessed using percentage agreement. Fidelity was assessed by examining the proportion of BCTs specified in the manuals that were delivered in individual sessions. This was assessed by session type (i.e., pre-quit, quit, post-quit), duration, individual practitioner, and BCT. Inter-coder reliability was high (87.1%). On average, 66% of manual-specified BCTs were delivered per session (SD 15.3, range: 35% to 90%). In Service 1, average fidelity was highest for post-quit sessions (69%) and lowest for pre-quit (58%). In Service 2, fidelity was highest for quit-day (81%) and lowest for post-quit sessions (56%). Session duration was not significantly correlated with fidelity. Individual practitioner fidelity ranged from 55% to 78%. Individual manual-specified BCTs were delivered on average 63% of the time (SD 28.5, range: 0 to 100%). The extent to which smoking cessation behavioural support is delivered as specified in treatment manuals can be reliably assessed using transcripts of audiotaped sessions. This allows the investigation of

Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

2010-07-01

SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

Rotary kiln and batch pyrolysis of waste tire to produce gasoline and diesel like fuels

International Nuclear Information System (INIS)

Ayanoğlu, Abdulkadir; Yumrutaş, Recep

2016-01-01

Highlights: • Waste Tire Oil (WTO) is produced from waste tire at rotary kiln reactor. • Physical and chemical properties of WTO and fuel samples are analyzed. • Gasoline like fuel (GLF) and diesel like fuel (DLF) are produced from the WTO-10 wt% CaO mixture at fixed bed reactor. • Physical and chemical properties of the GLF and DLF are compared with the standard fuels. - Abstract: In this study, waste tire is pyrolyzed in a rotary kiln reactor to obtain more gas, light liquid, heavy liquid, wax products, and less carbon black at their maximum yields as, 20%, 12%, 25%, 8% and 35% of the total weight (4 tones), respectively. Then, the heavy and light oils are reacted with additives such as natural zeolite (NZ) and lime (CaO) at different mass ratio as 2, 6, and 10 wt%, respectively, in the batch reactor to produce liquids similar to standard petroleum fuels. The heavy and light oils mixture samples are distillated to observe their optimum graphics which are similar to gasoline and diesel like fuel. Consequently, the best results are obtained from the CaO sample with 10 wt% in comparison to the ones from the gasoline and diesel fuels. The 10 wt% CaO light liquid mixture resembles to gasoline named as gasoline like fuel (GLF) and the 10 wt% CaO heavy liquid mixture is similar to diesel called as diesel like fuel (DLF). The chemical and physical features of the waste tire, light oil, heavy oil, GLF, and DLF are analyzed by TG (thermogravimetric)/dTG (derivative thermogravimetric), proximate, ultimate, higher heating value (HHV), fourier transform-infrared spectroscopy (FT-IR), Brunauer–Emmett–Teller (BET), sulfur, density, viscosity, gas chromatography–mass spectroscopy (GC–MS), flash point, moisture, and distillation tests. The test results are turned out to be very close to the standard petroleum fuel.

Experimental study of the thermal stability of hydrocarbon fuels

Science.gov (United States)

Marteney, P. J.; Colket, M. B.; Vranos, A.

1982-01-01

The thermal stability of two hydrocarbon fuels (premium diesel and regular diesel) was determined in a flow reactor under conditions representing operation of an aircraft gas turbine engine. Temperature was varied from 300 to 750 F (422 to 672 K) for fuel flows of 2.84 to 56.8 liters/hr (corresponding to 6.84 x 0.00010 to 1.63 x 0.010 kg/sec for regular diesel fuel and 6.55 x 0.00010 to 1.37 x 0.010 kg/sec for premium diesel fuel); test times varied between 1 and 8 hr. The rate of deposition was obtained through measurement of weight gained by metal discs fixed along the channel wall. The rate of deposit formation is best correlated by an Arrhenius expression. The sample discs in the flow reactor were varied among stainless steel, aluminum and brass; fuels were doped with quinoline, indole, and benzoyl perioxide to yield nitrogen or oxygen concentrations of approximately 1000 ppm. The most substantial change in rate was an increase in deposits for brass discs; other disc materials or the additives caused only small perturbations. Tests were also conducted in a static reactor at temperatures of 300 to 800 F for times of 30 min to 2 1/2 hr. Much smaller deposition was found, indicating the importance of fluid transport in the mechanism.

3D COMSOL Simulations for Thermal Deflection of HFIR Fuel Plate in the "Cheverton-Kelley" Experiments

Energy Technology Data Exchange (ETDEWEB)

Jain, Prashant K [ORNL; Freels, James D [ORNL; Cook, David Howard [ORNL

2012-08-01

Three dimensional simulation capabilities are currently being developed at Oak Ridge National Laboratory using COMSOL Multiphysics, a finite element modeling software, to investigate thermal expansion of High Flux Isotope Reactor (HFIR) s low enriched uranium fuel plates. To validate simulations, 3D models have also been developed for the experimental setup used by Cheverton and Kelley in 1968 to investigate the buckling and thermal deflections of HFIR s highly enriched uranium fuel plates. Results for several simulations are presented in this report, and comparisons with the experimental data are provided when data are available. A close agreement between the simulation results and experimental findings demonstrates that the COMSOL simulations are able to capture the thermal expansion physics accurately and that COMSOL could be deployed as a predictive tool for more advanced computations at realistic HFIR conditions to study temperature-induced fuel plate deflection behavior.

Transportation fuel production by combination of LDPE thermal cracking and catalytic hydroreforming

Energy Technology Data Exchange (ETDEWEB)

Escola, J.M., E-mail: josemaria.escola.saez@urjc.es [Department of Chemical and Environmental Technology, Rey Juan Carlos University, c/Tulipán s/n, 28933 Móstoles, Madrid (Spain); Aguado, J. [Department of Chemical and Environmental Technology, Rey Juan Carlos University, c/Tulipán s/n, 28933 Móstoles, Madrid (Spain); Serrano, D.P. [Department of Chemical and Environmental Technology, Rey Juan Carlos University, c/Tulipán s/n, 28933 Móstoles, Madrid (Spain); Department of Chemical and Energy Technology, Rey Juan Carlos University, c/Tulipán s/n, 28933 Móstoles, Madrid (Spain); IMDEA Energía, Avda. Ramón de la Sagra, 3, 28935 Móstoles, Madrid (Spain); Briones, L. [Department of Chemical and Environmental Technology, Rey Juan Carlos University, c/Tulipán s/n, 28933 Móstoles, Madrid (Spain)

2014-11-15

Highlights: • h-Beta samples were impregnated with Ni nitrate to achieve Ni contents of 1.5%, 4%, 7% and 10%. • Larger and more easily reducible metal particles were obtained on Ni 7%/h-Beta and Ni 10%/h-Beta. • Higher Ni contents increased the amount of gases at the expenses of diesel fractions. • Maximum selectivity to automotive fuels (∼81%) was obtained with Ni 7%/h-Beta. • Ni loading also enhanced olefins saturation up to Ni 7%/h-Beta. - Abstract: Fuel production from plastics is a promising way to reduce landfilling rates while obtaining valuable products. The usage of Ni-supported hierarchical Beta zeolite (h-Beta) for the hydroreforming of the oils coming from LDPE thermal cracking has proved to produce high selectivities to gasoline and diesel fuels (>80%). In the present work, the effect of the Ni loading on Ni/h-Beta is investigated in the hydroreforming of the oils form LDPE thermal cracking. h-Beta samples were impregnated with Ni nitrate, calcined and reduced in H{sub 2} up to 550 °C to achieve different Ni contents: 1.5%, 4%, 7% and 10%. Larger and more easily reducible metal particles were obtained on Ni 7%/h-Beta and Ni 10%/h-Beta. Hydroreforming tests were carried out in autoclave reactor at 310 °C, under 20 bar H{sub 2}, for 45 min. Ni content progressively increased the amount of gases at the expenses of diesel fractions, while gasoline remained approximately constant about 52–54%. Maximum selectivity to automotive fuels (∼81%) was obtained with Ni 7%/h-Beta. Ni loading also enhanced olefins saturation up to Ni 7%/h-Beta. High cetane indices (71–86) and octane numbers (89–91) were obtained over all the catalysts. Regarding the different studied Ni contents, Ni 7%/h-Beta constitutes a rather promising catalyst for obtaining high quality fuels from LDPE thermal cracking oils.

Transportation fuel production by combination of LDPE thermal cracking and catalytic hydroreforming

International Nuclear Information System (INIS)

Escola, J.M.; Aguado, J.; Serrano, D.P.; Briones, L.

2014-01-01

Highlights: • h-Beta samples were impregnated with Ni nitrate to achieve Ni contents of 1.5%, 4%, 7% and 10%. • Larger and more easily reducible metal particles were obtained on Ni 7%/h-Beta and Ni 10%/h-Beta. • Higher Ni contents increased the amount of gases at the expenses of diesel fractions. • Maximum selectivity to automotive fuels (∼81%) was obtained with Ni 7%/h-Beta. • Ni loading also enhanced olefins saturation up to Ni 7%/h-Beta. - Abstract: Fuel production from plastics is a promising way to reduce landfilling rates while obtaining valuable products. The usage of Ni-supported hierarchical Beta zeolite (h-Beta) for the hydroreforming of the oils coming from LDPE thermal cracking has proved to produce high selectivities to gasoline and diesel fuels (>80%). In the present work, the effect of the Ni loading on Ni/h-Beta is investigated in the hydroreforming of the oils form LDPE thermal cracking. h-Beta samples were impregnated with Ni nitrate, calcined and reduced in H 2 up to 550 °C to achieve different Ni contents: 1.5%, 4%, 7% and 10%. Larger and more easily reducible metal particles were obtained on Ni 7%/h-Beta and Ni 10%/h-Beta. Hydroreforming tests were carried out in autoclave reactor at 310 °C, under 20 bar H 2 , for 45 min. Ni content progressively increased the amount of gases at the expenses of diesel fractions, while gasoline remained approximately constant about 52–54%. Maximum selectivity to automotive fuels (∼81%) was obtained with Ni 7%/h-Beta. Ni loading also enhanced olefins saturation up to Ni 7%/h-Beta. High cetane indices (71–86) and octane numbers (89–91) were obtained over all the catalysts. Regarding the different studied Ni contents, Ni 7%/h-Beta constitutes a rather promising catalyst for obtaining high quality fuels from LDPE thermal cracking oils

Internal Nozzle Flow Simulations of Gasoline-Like Fuels under Diesel Operating Conditions

Energy Technology Data Exchange (ETDEWEB)

Torelli, R.; Som, S.; Pei, Y.; Zhang, Yu; Traver, Michael

2017-05-15

Spray formation in internal combustion engines with direct injection is strictly correlated with internal nozzle flow characteristics, which are in turn influenced by fuel physical properties and injector needle motion. This paper pre-sents a series of 3D simulations that model the in-nozzle flow in a 5-hole mini-sac diesel injector. Two gasoline-like naphtha fuels, namely full-range and light naphtha, were tested under operating conditions typical of diesel applica-tions and were compared with n-dodecane, selected from a palette used as diesel surrogates. Validated methodolo-gies from our previous work were employed to account for realistic needle motion. The multi-phase nature of the problem was described by the mixture model assumption with the Volume of Fluid method. Cavitation effects were included by means of the Homogeneous Relaxation Model and turbulence closure was achieved with the Standard k-ε model in an Unsteady Reynolds-Averaged Navier-Stokes formulation. The results revealed that injector perfor-mance and propensity to cavitation are influenced by the fuel properties. Analyses of several physical quantities were carried out to highlight the fuel-to-fuel differences in terms of mass flow rate, discharge coefficients, and fuel vapor volume fraction inside the orifices. A series of parametric investigations was also performed to assess the fuel response to varied fuel injection temperature, injection pressure, and cross-sectional orifice area. For all cases, the strict correlation between cavitation magnitude and saturation pressure was confirmed. Owing to their higher volatil-ity, the two gasoline-like fuels were characterized by higher cavitation across all the simulated conditions. Occur-rence of cavitation was mostly found at the needle seat and at the orifice inlets during the injection event’s transient, when very small gaps exist between the needle and its seat. This behavior tended to disappear at maximum needle lift, where cavitation was

Comparison of thermal capabilities of the fuel assemblies for the WWR-M reactor

International Nuclear Information System (INIS)

Kirsanov, G.A.; Konoplev, K.A.; Findeisen, A.; Shishkina, Zh.A.

1989-01-01

On the basis of measurement results of the WWR-M2, WWR-M3 and WWR-M5 fuel element can temperature in the WWR-M reactor core their thermal capabilities are compared. The use of the WWR-M5 fuel assemblies instead of the WWR-M2 ones in the WWR-M reactor permits to increase specific heat loading by a factor of 2.7. The possibility to increase fuel can temperature up to 110 deg C is confirmed experimentally which corresponds to specific heat loading of 900 kW/l

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

International Nuclear Information System (INIS)

Unterzuber, R.; Hanson, J.P.

1981-09-01

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

Mosaic-shaped cathode for highly durable solid oxide fuel cell under thermal stress

Science.gov (United States)

Joo, Jong Hoon; Jeong, Jaewon; Kim, Se Young; Yoo, Chung-Yul; Jung, Doh Won; Park, Hee Jung; Kwak, Chan; Yu, Ji Haeng

2014-02-01

In this study, we propose a novel "mosaic structure" for a SOFC (solid oxide fuel cell) cathode with high thermal expansion to improve the stability against thermal stress. Self-organizing mosaic-shaped cathode has been successfully achieved by controlling the amount of binder in the dip-coating solution. The anode-supported cell with mosaic-shaped cathode shows itself to be highly durable performance for rapid thermal cycles, however, the performance of the cell with a non-mosaic cathode exhibits severe deterioration originated from the delamination at the cathode/electrolyte interface after 7 thermal cycles. The thermal stability of an SOFC cathode can be evidently improved by controlling the surface morphology. In view of the importance of the thermal expansion properties of the cathode, the effects of cathode morphology on the thermal stress stability are discussed.

Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

2015-12-31

A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

The thermal conductivity of mixed fuel U_xPu_1-xO₂: molecular dynamics simulations

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cooper, Michael William Donald [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-10-16

Mixed oxides (MOX), in the context of nuclear fuels, are a mixture of the oxides of heavy actinide elements such as uranium, plutonium and thorium. The interest in the UO₂-PuO₂ system arises from the fact that these oxides are used both in fast breeder reactors (FBRs) as well as in pressurized water reactors (PWRs). The thermal conductivity of UO₂ fuel is an important material property that affects fuel performance since it is the key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. For this reason it is important to understand the thermal conductivity of MOX fuel and how it differs from UO₂. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of mixing on the thermal conductivity of U_xPu_1-xO₂, as a function of PuO₂ concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel.

Effect of burnup on the response of stainless steel-clad mixed-oxide fuels to simulated thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Badyopadhyay, G.

1981-01-01

Direct electrical heating experiments were performed on irradiated fuel to study the fuel and cladding response as a function of burnup during a slow thermal transient. The results indicated that the nature and extent of the fuel and cladding behavior depended on the quantity of fission gas retained in the fuel. Fission-gas-driven fuel ejection occurred as the molten cladding flowed down the stack exposing bare, radially unrestrained fuel. The fuel dispersion occurred in the absence of molten fuel and the amount of fuel ejected increased with increasing burnup. 31 refs

Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

International Nuclear Information System (INIS)

Medvedev, Pavel G.

2009-01-01

Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20 C temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.

A fuel thermal conductivity correlation based on the latest experimental results

International Nuclear Information System (INIS)

Sontheimer, F.; Landskron, H.; Billaux, M.R.

2000-01-01

A new fuel thermal conductivity (ftc) correlation for UO 2 and (U,Gd)O 2 is presented, which is based on the relaxation-time theory of Klemens. The correlation is chosen because of its validity in a wide range of defect concentrations as for instance encountered in fuel with a wide range of burnup and gadolinia additions, as has been shown by Ishimoto. The phonon term of the new correlation has the form 1/x·arctan(x) , where x is a measure of the defect concentration introduced by burnup and gadolinia additions. For low defect concentrations, this term is identical with the classical form for the phonon term 1/(A+B.T). At high defect concentrations, however, when phonon-point defect scattering starts dominating over phonon-phonon scattering, the new correlation deviates from the classical formulation and has a distinctly weaker dependence on temperature and defect concentration than the classical form. The new arctan correlation in combination with an appropriate electronic ftc term is fitted to the Halden data base of fuel centre-line temperature measurements (represented by the ''Halden ftc correlation recommendation''). Agreement is very good up to a burnup of about 60 MWd/kgU; beyond, the arctan form has a saturating burnup degradation. The new arctan correlation in combination with an appropriate electronic ftc term is also shown to describe very well our latest ftc measurements on unirradiated gadolinia fuel up to 9% gadolinia content. Application to Halden measurements up to very high burnup is successful, when combined with the so-called ''rim-effect'', which counteracts the saturation tendency of the new correlation at high burnup. Latest laser thermal diffusivity measurements on irradiated gadolinia fuel in the frame of the NFIR program, although not yet open for literature and not discussed in the paper, indicate very good agreement with the new arctan correlation. (author)

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)

Thermohydraulic and thermal stress aspects of a porous blockage in an LMFBR fuel assembly

International Nuclear Information System (INIS)

Kuzay, T.M.; Marr, W.W.; Helenberg, H.W.; Ariman, T.; Wilson, R.E.; Pedersen, D.R.

1979-01-01

The current safety scenarios of Liquid Metal Fast Breeder Reactors (LMFBR) under local fault propagation include the study of a hypothetical accident initiated by the formation of an external debris porous blockage in a fuel subassembly. In this preliminary experimental and analytical investigation, a non-heat-generating porous blockage was postulated to cover 18 flow channels of a 37 pin Fast Test Reactor (FTR) type fuel subassembly. The axial extent of the blockage is 50 mm. The blockage material is stainless steel (SS 316) with 30 percent average porosity (percent void volume). The blockage and the pins were modeled with a finite element technique and the thermal field in the blockage was predicted. This thermal field was utilized to do a planar thermal stress analysis of the postulated blockage. To verify the analytical model and also to better understand the thermal-hydraulics of such a porous blockage out-of-pile tests were conducted in a sodium loop. Data from the out-of-pile tests was utilized to calibrate and improve the analytical model

Development of MHI PWR fuel assembly with high thermal performance

International Nuclear Information System (INIS)

Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

2005-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

1981-01-01

The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

Fiscal loss and program fidelity: impact of the economic downturn on HIV/STI prevention program fidelity.

Science.gov (United States)

Catania, Joseph A; Dolcini, M Margaret; Gandelman, Alice A; Narayanan, Vasudha; McKay, Virginia R

2014-03-01

The economic downturn of 2007 created significant fiscal losses for public and private agencies conducting behavioral prevention. Such macro-economic changes may influence program implementation and sustainability. We examined how public and private agencies conducting RESPECT, a brief HIV/STI (sexually transmitted infection) counseling and testing intervention, adapted to fiscal loss and how these adaptations impacted program fidelity. We collected qualitative and quantitative data in a national sample of 15 agencies experiencing fiscal loss. Using qualitative analyses, we examined how program fidelity varied with different types of adaptations. Agencies reported three levels of adaptation: agency-level, program-level, and direct fiscal remedies. Private agencies tended to use direct fiscal remedies, which were associated with higher fidelity. Some agency-level adaptations contributed to reductions in procedural fit, leading to negative staff morale and decreased confidence in program effectiveness, which in turn, contributed to poor fidelity. Findings describe a "work stress pathway" that links program fiscal losses to poor staff morale and low program fidelity.

Study on Thermal Degradation Characteristics and Regression Rate Measurement of Paraffin-Based Fuel

Directory of Open Access Journals (Sweden)

Songqi Hu

2015-09-01

Full Text Available Paraffin fuel has been found to have a regression rate that is higher than conventional HTPB (hydroxyl-terminated polybutadiene fuel and, thus, presents itself as an ideal energy source for a hybrid rocket engine. The energy characteristics of paraffin-based fuel and HTPB fuel have been calculated by the method of minimum free energy. The thermal degradation characteristics were measured for paraffin, pretreated paraffin, HTPB and paraffin-based fuel in different working conditions by the using differential scanning calorimetry (DSC and a thermogravimetric analyzer (TGA. The regression rates of paraffin-based fuel and HTPB fuel were tested by a rectangular solid-gas hybrid engine. The research findings showed that: the specific impulse of paraffin-based fuel is almost the same as that of HTPB fuel; the decomposition temperature of pretreated paraffin is higher than that of the unprocessed paraffin, but lower than that of HTPB; with the increase of paraffin, the initial reaction exothermic peak of paraffin-based fuel is reached in advance, and the initial reaction heat release also increases; the regression rate of paraffin-based fuel is higher than the common HTPB fuel under the same conditions; with the increase of oxidizer mass flow rate, the regression rate of solid fuel increases accordingly for the same fuel formulation.

Fuel assembly

International Nuclear Information System (INIS)

Nomata, Terumitsu.

1993-01-01

Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

Comparison of thermal compatibility between atomized and comminuted U{sub 3}Si dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu; Kuk, II-Hyun [Korea Atomic Research Institute, Taejon (Korea, Republic of)

1997-08-01

Thermal compatibility of atomized U{sub 3}Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500{degrees}C, and compared with that of comminuted U{sub 3}Si. Atomized U{sub 3}Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U{sub 3}Si and Al occurred along the grain boundaries and deformation bands in U{sub 3}Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, and reduced the volume expansion.

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

Science.gov (United States)

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

2011-01-01

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

Thermal Expansion Property of U-Zr Alloys and U-Zr-Ce Alloys as a Surrogate Metallic Fuel for SFR

International Nuclear Information System (INIS)

Kim, Sun Ki; Lee, Jong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Youn Myung; Lee, Chan Bock

2010-01-01

Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. An extensive database on the performance of advanced metal fuels was generated as a result of the operation of these reactors and the IFR program. In this study, the U-Zr binary alloys and U-Zr-Ce ternary alloys as surrogate metallic fuel were fabricated in lower pressure Ar environment by gravity casting. The melt temperature was approximately 1,500 .deg. C. Thermal expansion of the fuel during normal operation is related with fuel performance in a reactor. Therefore, it is necessary to investigate the thermal expansion of the fuel in order to warrant a good prediction the fuel performance

Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

Energy Technology Data Exchange (ETDEWEB)

Xu, Yumei [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Yang, Jian, E-mail: zdhjkz@zju.edu.cn [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Xu, Chao; Wang, Weiping [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Ma, Zhijun [Department of Material Engineering, South China University of Technology, Guangzhou (China)

2013-12-15

Highlights: • Spent fuel cask was investigated as a whole instead of fuel assembly alone. • The cask was successfully modeled and meshed after several simplifications. • Equivalence method was used to calculate the properties of parts. • Both the integral thermal field and peak values are captured to verify safety. • The temperature variations of key parts were also plotted. - Abstract: Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

Fidelity in clinical simulation

DEFF Research Database (Denmark)

Jensen, Sanne; Nøhr, Christian; Rasmussen, Stine Loft

2013-01-01

Clinical simulation may be used to identify user needs for context sensitive functionalities in e-Health. The objective with this paper is to describe how user requirements and use cases in a large EHR-platform procurement may be validated by clinical simulation using a very low-fidelity prototype...... without any existing test data. Instead of using test scenarios and use cases, the healthcare professionals who are participating in the clinical simulation are generating both scenario and patient data themselves. We found that this approach allows for an imaginative discussion, not restricted by known...... functionalities and limitations, of the ideal EHR-platform. Subsequently, we discuss benefits and challenges of using an extremely low fidelity environment and discuss the degree of fidelity necessary for conducting clinical simulation....

Thermal Conductivity of Metallic Uranium

Energy Technology Data Exchange (ETDEWEB)

Hin, Celine

2018-03-10

This project has developed a modeling and simulation approaches to predict the thermal conductivity of metallic fuels and their alloys. We focus on two methods. The first method has been developed by the team at the University of Wisconsin Madison. They developed a practical and general modeling approach for thermal conductivity of metals and metal alloys that integrates ab-initio and semi-empirical physics-based models to maximize the strengths of both techniques. The second method has been developed by the team at Virginia Tech. This approach consists of a determining the thermal conductivity using only ab-initio methods without any fitting parameters. Both methods were complementary. The models incorporated both phonon and electron contributions. Good agreement with experimental data over a wide temperature range were found. The models also provided insight into the different physical factors that govern the thermal conductivity under different temperatures. The models were general enough to incorporate more complex effects like additional alloying species, defects, transmutation products and noble gas bubbles to predict the behavior of complex metallic alloys like U-alloy fuel systems under burnup. 3 Introduction Thermal conductivity is an important thermal physical property affecting the performance and efficiency of metallic fuels [1]. Some experimental measurement of thermal conductivity and its correlation with composition and temperature from empirical fitting are available for U, Zr and their alloys with Pu and other minor actinides. However, as reviewed in by Kim, Cho and Sohn [2], due to the difficulty in doing experiments on actinide materials, thermal conductivities of metallic fuels have only been measured at limited alloy compositions and temperatures, some of them even being negative and unphysical. Furthermore, the correlations developed so far are empirical in nature and may not be accurate when used for prediction at conditions far from those

Nonlinear analysis and evaluation of a reinforced concrete spent fuel storage pool for accidental thermal loads

International Nuclear Information System (INIS)

Kabir, A.F.; Bolourchi, S.

1991-01-01

A feasibility study was conducted for addition of consolidated fuel racks to an existing reinforced concrete spent fuel storage pool of a Mark I BWR plant. Nonlinear analysis of a detailed three-dimensional model of the fuel pool, considering cracking in concrete under gravity and thermal load conditions, showed that the pool has reserve capacities to carry the additional loads. (author)

Fuel combustion in thermal power plants in Japan

Energy Technology Data Exchange (ETDEWEB)

Kotler, V.R.

1983-11-01

The position of black coal in the energy balance of Japan is discussed. About 75% of electric energy is produced by thermal power plants. Eighty-five per cent of electricity is produced by power plants fired with liquid fuels and 3% by coal fired plants. Coal production in Japan, the forecast coal import to the country by 1990 (132 Mt/year), proportion of coal imported from various countries, chemical and physical properties of coal from Australia, China and Japan are discussed. Coal classification used in Japan is evaluated. The following topics associated with coal combustion in fossil-fuel power plants in Japan are discussed: coal grindability, types of pulverizing systems, slagging properties of boiler fuel in Japan, systems for slag removal, main types of steam boilers and coal fired furnaces, burner arrangement and design, air pollution control from fly ash, sulfur oxides and nitrogen oxides, utilization of fly ash for cement production, methods for removal of nitrogen oxides from flue gas using ammonia and catalysts or ammonia without catalysts, efficiency of nitrogen oxide control, abatement of nitrogen oxide emission from boilers by flue gas recirculation and reducing combustion temperatures. The results of research into air pollution control carried out by the Nagasaki Technical Institute are reviewed.

Non-linear thermal and structural analysis of a typical spent fuel silo

International Nuclear Information System (INIS)

Alvarez, L.M.; Mancini, G.R.; Spina, O.A.F.; Sala, G.; Paglia, F.

1993-01-01

A numerical method for the non-linear structural analysis of a typical reinforced concrete spent fuel silo under thermal loads is proposed. The numerical time integration was performed by means of a time explicit axisymmetric finite-difference numerical operator. An analysis was made of influences by heat, viscoelasticity and cracking upon the concrete behaviour between concrete pouring stage and the first period of the silo's normal operation. The following parameters were considered for the heat generation and transmission process: Heat generated during the concrete's hardening stage, Solar radiation effects, Natural convection, Spent-fuel heat generation. For the modelling of the reinforced concrete behaviour, use was made of a simplified formulation of: Visco-elastic effects, Thermal cracking, Steel reinforcement. A comparison between some experimental temperature characteristic values obtained from the numerical integration process and empirical data obtained from a 1:1 scaled prototype was also carried out. (author)

Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

International Nuclear Information System (INIS)

Thakre, Sachin; Lakshmanan, S.P.; Kulkarni, Vinayak; Pandey, Manmohan

2009-01-01

Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

Thermal Expansion and Density Data of UO2 and Simulated Fuel for Standard Reference

International Nuclear Information System (INIS)

Yang, Jae Hwan; Na, S. H.; Lee, J. W.; Kang, K. H.

2010-01-01

Standard Reference Data (SRD) is the scientific, technical data whose reliability and accuracy are evaluated by scientist group. Since SRD has a great impact on the improvement of national competitiveness by stirring up technological innovation in every sector of industries, many countries are making great efforts on establishing SRD in various areas. Data center for nuclear fuel material in Korea Atomic Energy Research Institute plays a role to providing property data of nuclear fuel material at high temperature, pressure, and radiation which are essential for the safety evaluation of nuclear power. In this study, standardization of data on thermal expansion and density of UO 2 were carried out in the temperature range from 300 K to 3100 K via uncertainty evaluation of indirectly produced data. Besides, standardization of data on thermal expansion and density of simulated fuel were also done in the temperature range from 350 K to 1750 K via uncertainty evaluation of directly produced data

Liquid-like thermal conduction in intercalated layered crystalline solids

Science.gov (United States)

Li, B.; Wang, H.; Kawakita, Y.; Zhang, Q.; Feygenson, M.; Yu, H. L.; Wu, D.; Ohara, K.; Kikuchi, T.; Shibata, K.; Yamada, T.; Ning, X. K.; Chen, Y.; He, J. Q.; Vaknin, D.; Wu, R. Q.; Nakajima, K.; Kanatzidis, M. G.

2018-03-01

As a generic property, all substances transfer heat through microscopic collisions of constituent particles1. A solid conducts heat through both transverse and longitudinal acoustic phonons, but a liquid employs only longitudinal vibrations2,3. As a result, a solid is usually thermally more conductive than a liquid. In canonical viewpoints, such a difference also serves as the dynamic signature distinguishing a solid from a liquid. Here, we report liquid-like thermal conduction observed in the crystalline AgCrSe2. The transverse acoustic phonons are completely suppressed by the ultrafast dynamic disorder while the longitudinal acoustic phonons are strongly scattered but survive, and are thus responsible for the intrinsically ultralow thermal conductivity. This scenario is applicable to a wide variety of layered compounds with heavy intercalants in the van der Waals gaps, manifesting a broad implication on suppressing thermal conduction. These microscopic insights might reshape the fundamental understanding on thermal transport properties of matter and open up a general opportunity to optimize performances of thermoelectrics.

Three-dimensional FE analysis of the thermal-mechanical behaviors in the nuclear fuel rods

International Nuclear Information System (INIS)

Jiang Yijie; Cui Yi; Huo Yongzhong; Ding Shurong

2011-01-01

Highlights: → We establish three-dimensional finite element models for nuclear fuel rods. → The thermal-mechanical behaviors at the initial stage of burnup are obtained. → Several parameters on the in-pile performances are investigated. → The parameters have remarkable effects on the in-pile behaviors. → This study lays a foundation for optimal design and irradiation safety. - Abstract: In order to implement numerical simulation of the thermal-mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal-mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal-mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm 3 , the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm 2 K to 0.01 W/mm 2 K, while increases up to 54.7% when h decreases from 0.01 W/mm 2 K to 0.005 W/mm 2 K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal-mechanical behaviors in the fuel rod; when the

FEMAXI-III: a computer code for the analysis of thermal and mechanical behavior of fuel rods

International Nuclear Information System (INIS)

Nakajima, Tetsuo; Ichikawa, Michio; Iwano, Yoshihiko; Ito, Kenichi; Saito, Hiroaki; Kashima, Koichi; Kinoshita, Motoyasu; Okubo, Tadatsune.

1985-12-01

FEMAXI-III is a computer code to predict the thermal and mechanical behavior of a light water fuel rod during its irradiation life. It can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized behavior of a small part of fuel rod. The localized mechanical behavior such as the cladding ridge deformation is analyzed by the two-dimensional axisymmetric finite element method. FEMAXI-III calculates, in particular, the temperature distribution, the radial deformation, the fission gas release, and the inner gas pressure as a function of irradiation time and axial position, and the stresses and strains in the fuel and cladding at a small part of fuel rod as a function of irradiation time. For this purpose, Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled and their interconnected effects are considered in the code. Efforts have been made to improve the accuracy and stability of finite element solution and to minimize the computer memory and running time. This report describes the outline of the code and the basic models involved, and also includes the application of the code and its input manual. (author)

Role of thermal analysis in uranium oxide fuel fabrication process

International Nuclear Information System (INIS)

Balaji Rao, Y.; Yadav, R.B.

2006-01-01

The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO 2+x and U 3 O 8 powders as an alternate technique. Further, the temperature difference (ΔT max ) between the two exothermic peaks obtained in UO 2+x powder oxidation is related to sintered density of UO 2 pellets. (author)

SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements

International Nuclear Information System (INIS)

Hayes, S.L.

1993-12-01

SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user's manual. A sample calculation made with SAFE is included to highlight some of the code's features. Complete input and output files for the sample problem are provided

James Webb Space Telescope Core 2 Test - Cryogenic Thermal Balance Test of the Observatorys Core Area Thermal Control Hardware

Science.gov (United States)

Cleveland, Paul; Parrish, Keith; Thomson, Shaun; Marsh, James; Comber, Brian

2016-01-01

The James Webb Space Telescope (JWST), successor to the Hubble Space Telescope, will be the largest astronomical telescope ever sent into space. To observe the very first light of the early universe, JWST requires a large deployed 6.5-meter primary mirror cryogenically cooled to less than 50 Kelvin. Three scientific instruments are further cooled via a large radiator system to less than 40 Kelvin. A fourth scientific instrument is cooled to less than 7 Kelvin using a combination pulse-tube Joule-Thomson mechanical cooler. Passive cryogenic cooling enables the large scale of the telescope which must be highly folded for launch on an Ariane 5 launch vehicle and deployed once on orbit during its journey to the second Earth-Sun Lagrange point. Passive cooling of the observatory is enabled by the deployment of a large tennis court sized five layer Sunshield combined with the use of a network of high efficiency radiators. A high purity aluminum heat strap system connects the three instrument's detector systems to the radiator systems to dissipate less than a single watt of parasitic and instrument dissipated heat. JWST's large scale features, while enabling passive cooling, also prevent the typical flight configuration fully-deployed thermal balance test that is the keystone of most space missions' thermal verification plans. This paper describes the JWST Core 2 Test, which is a cryogenic thermal balance test of a full size, high fidelity engineering model of the Observatory's 'Core' area thermal control hardware. The 'Core' area is the key mechanical and cryogenic interface area between all Observatory elements. The 'Core' area thermal control hardware allows for temperature transition of 300K to approximately 50 K by attenuating heat from the room temperature IEC (instrument electronics) and the Spacecraft Bus. Since the flight hardware is not available for test, the Core 2 test uses high fidelity and flight-like reproductions.

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

International Nuclear Information System (INIS)

Niceno, Bojan; Pouchon, Manuel

2014-01-01

The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

Development of uranium dioxide fuel pellets with addition of beryllium oxide for increasing of thermal conductivity

International Nuclear Information System (INIS)

Queiroz, Carolinne Mol; Ferreira, Ricardo Alberto Neto

2011-01-01

The CDTN - Centro de Desenvolvimento de Tecnologia Nuclear presents a project named 'Beryllium Project' viewing to increasing the thermal conductivity of UO 2 fuel pellets, increasing the lifetime of those pellets in the reactor, generating a greater economy. This increase of conductivity is obtained by means of Be O addition to the UO 2 fuel pellets, which is very used for the production of nuclear energy. The UO 2 pellets however present a thermal conductivity relatively low, generating a high temperature gradient between the center and his side surface. The addition of beryllium oxide, with higher thermal conductivity gives pellets which will present lower temperature gradient and, consequently, more durability and better utilization of energy potential of the pellet in the reactor. (author)

Plutonium rock-like fuel LWR nuclear characteristics and transient behavior in accidents

Energy Technology Data Exchange (ETDEWEB)

Akie, Hiroshi; Anoda, Yoshinari; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yamaguchi, Chouichi; Sugo, Yukihiro

1998-03-01

For the disposition of excess plutonium, rock-like oxide (ROX) fuel systems based on zirconia (ZrO{sub 2}) or thoria (ThO{sub 2}) have been studied. Safety analysis of ROX fueled PWR showed it is necessary to increase Doppler reactivity coefficient and to reduce power peaking factor of zirconia type ROX (Zr-ROX) fueled core. For these improvements, Zr-ROX fuel composition was modified by considering additives of ThO{sub 2}, UO{sub 2} or Er{sub 2}O{sub 3}, and reducing Gd{sub 2}O{sub 3} content. As a result of the modification, comparable, transient behavior to UO{sub 2} fuel PWR was obtained with UO{sub 2}-Er{sub 2}O{sub 3} added Zr-ROX fuel, while the plutonium transmutation capability is slightly reduced. (author)

Numerical simulation of ambient flow and thermal distributions in a spent fuel storage cask array

International Nuclear Information System (INIS)

Michener, T.; Trent, D.S.; Guttmann, J.; Bajwa, C.

2001-01-01

At the request of the U.S. Nuclear Regulatory Commission (USNRC), the staff at the Pacific Northwest National Laboratory (PNNL) analyzed the thermal performance of the Utah Private Fuel Storage (PFS) using the TEMPEST computational fluid dynamics software. A three-dimensional section of the PFS with a total of 20 casks was modeled to estimate the ambient flow and temperature distributions surrounding the casks. The purpose of this analysis was to compute the cask inlet vent air temperature to be used for boundary conditions in a detailed analysis of an individual Holtec Hi-Storm 100 cask using the COBRA-SFS (Spent Fuel Storage) thermal hydraulic computer software. (author)

Evaluation of parameters effect on the maximum fuel temperature in the core thermal and hydraulic design of HTTR

International Nuclear Information System (INIS)

Fujimoto, Nozomu; Maruyama, Soh; Sudo, Yukio; Fujii, Sadao; Niguma, Yoshinori.

1988-10-01

This report presents the results of quantitative evaluation on the effects of the dominant parameters on the maximum fuel temperature in the core thermal hydraulic design of the High Temperature Engineering Test Reactor(HTTR) of 30 MW in thermal power, 950 deg C in reactor outlet coolant temperature and 40 kg/cm 2 G in coolant pressure. The dominant parameters investigated are 1) Gap conductance. 2) Effect of eccertricity of fuel compacts in graphite sleeve. 3) Effect of spacer ribs on heat transfer coefficients. 4) Contact probability of fuel compact and graphite sleeve. 5) Validity of uniform radial power density in the fuel compacts. 6) Effect of impurity gas on gap conductance. 7) Effect of FP gas on gap conductance. The effects of these items on the maximum fuel temperature were quantitalively identified as hot spot factors. A probability of the appearance of the maximum fuel temperature was also evaluated in this report. (author)

Evaluation of Effective thermal conductivity models on the prismatic fuel block of a Very High Temperature Reactor by CFD analysis

International Nuclear Information System (INIS)

Shin, Dong-Ho; Cho, Hyoung-Kyu; Tak, Nam-Il; Park, Goon-Cherl

2014-01-01

Effective thermal conductivity models which can be used to analyze the heat transfer phenomena of a prismatic fuel block were evaluated by CFD analysis. In the accident condition of VHTR when forced convection is lost, the heat flows in radial direction through the hexagonal fuel blocks that contain the large number of coolant holes and fuel compacts. Due to the complex geometry of fuel block and radiation heat transfer; the detail heat transfer computation on the fuel block needs excessive computation resources. Therefore, the detail computation isn’t appropriate for the lumped parameter code. The system code such as GAMMA+ adopts effective thermal conductivity model. Despite the complexity in heat transfer modes, the accurate analysis on the heat transfer in fuel block is necessary since it is directly relevant to the integrity of nuclear fuel embedded in fuel block. To satisfy the accurate analysis of complex heat transfer modes with limited computing sources, the credible effective thermal conductivity (ETC) models in which the effects of all of heat transfer modes are lumped is necessary. In this study, various ETC models were introduced and they are evaluated with CFD calculations. It is estimated that Maxwell-based model was the most pertinent one among the introduced ETC models. (author)

Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

International Nuclear Information System (INIS)

Ohshima, Hiroyuki

2001-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

Heat transfer and fluid flow aspects of fuel--coolant interactions

International Nuclear Information System (INIS)

Corradini, M.L.

1978-09-01

A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

The fuel and channel thermal/mechanical behaviour code FACTAR 2.0 (LOCA)

International Nuclear Information System (INIS)

Westbye, C.J.; Mackinnon, J.C.; Gu, B.W.

1996-01-01

The computer code FACTAR 2.0 (LOCA) models the thermal and mechanical response of components within a single CANDU fuel channel under loss-of-coolant accident conditions. This code version is the successor to the FACTAR 1.x code series, and features many modelling enhancements over its predecessor. In particular, the thermal hydraulic treatment has been extended to model reverse and bi-directional coolant flow, and the axial variation in coolant flow rate. Thermal radiation is calculated by a detailed surface-to-surface model, and the ability to represent a greater range of geometries (including experimental configurations employed in code validation) has been implemented. Details of these new code treatments are described in this paper. (author)

A comprehensive review on the methodologies to simulate the nuclear fuel bundle for the thermal hydraulic experiments

International Nuclear Information System (INIS)

Vishnoi, A.K.; Chandraker, D.K.; Pal, A.K.; Vijayan, P.K.; Saha, D.

2011-01-01

The designer of a nuclear reactor system has to ensure its safety during normal operation as well as accidental conditions. This requires, among other things, a proper understanding of the various thermal hydraulic phenomena occurring in the reactor core. In a nuclear reactor core the fuel elements are the heat source and highly loaded components of the reactor system. Therefore their behaviour under normal and accidental conditions must be extensively investigated. Data generation for Critical heat flux (CHF) in full scale bundle and parallel channel instability studies with at least two full size channels are required in order to evaluate the thermal margin and stability margin of the reactor. The complex nature of these phenomena calls for exhaustive experimental investigations. Fuel Rod Cluster Simulator (FRCS) is a very important component required for the experimental investigation of the thermal hydraulic behaviour of reactor fuel elements under normal and accidental conditions. This paper brings out a comprehensive review of the FRCS elaborating the challenges and important design aspects of the FRCS. Some of the main features and analysis results on the performance of the developed FRCS with respect to the actual nuclear fuel bundle will be presented in the paper. (author)

Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

International Nuclear Information System (INIS)

Ethridge, J.L.; Baker, R.B.

1985-01-01

A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

Thermal conductivity of high performance carbon nanotube yarn-like fibers

Energy Technology Data Exchange (ETDEWEB)

Mayhew, Eric; Prakash, Vikas, E-mail: vikas.prakash@case.edu [Department of Mechanical and Aerospace Engineering, Case Western Reserve University, Cleveland, Ohio 44106-7222 (United States)

2014-05-07

In the present paper, we present results of thermal conductivity measurements in free standing carbon nanotube (CNT) yarn-like fibers. The measurements are made using a T-type experimental configuration utilizing a Wollaston-wire hot probe inside a scanning electron microscope. In this technique, a suspended platinum wire is used both as a heater and a thermal sensor. A low frequency alternating current source is used to heat the probe wire while the third harmonic voltage across the wire is measured by a lock-in amplifier. The conductivity is deduced from an analytical model that relates the drop in the spatially averaged temperature of the wire to that of the sample. The average thermal conductivity of the neat CNT fibers and the CNT –polymer composite fibers is found to be 448 W/m-K and 225 W/m-K, respectively. These values for conductivity are amongst the highest measured for CNT yarn-like fibers fabricated using a dry spinning process from vertically aligned CNT arrays. The enhancement in thermal conductivity is understood to be due to an increase in the CNT fiber elastic stiffness during the draw and twist operations, lower CNT thermal contact resistance due to increase in CNT contact area, and better alignment of the CNT fibrils along the length of the fiber.

High-Fidelity Simulation of Pediatric Emergency Care: An Eye-Opening Experience for Baccalaureate Nursing Students.

Science.gov (United States)

Small, Sandra P; Colbourne, Peggy A; Murray, Cynthia L

2018-01-01

Background Little attention has been given to in-depth examination of what high-fidelity simulation is like for nursing students within the context of a pediatric emergency, such as a cardiopulmonary arrest. It is possible that such high-fidelity simulation could provoke in nursing students intense psychological reactions. Purpose The purpose of this study was to learn about baccalaureate nursing students' lived experience of high-fidelity simulation of pediatric cardiopulmonary arrest. Method Phenomenological methods were used. Twenty-four interviews were conducted with 12 students and were analyzed for themes. Results The essence of the experience is that it was eye-opening. The students found the simulation to be a surprisingly realistic nursing experience as reflected in their perceiving the manikin as a real patient, thinking that they were saving their patient's life, feeling like a real nurse, and feeling relief after mounting stress. It was a surprisingly valuable learning experience in that the students had an increased awareness of the art and science of nursing and increased understanding of the importance of teamwork and were feeling more prepared for clinical practice and wanting more simulation experiences. Conclusion Educators should capitalize on the benefits of high-fidelity simulation as a pedagogy, while endeavoring to provide psychologically safe learning.

Simulation of thermal stresses in anode-supported solid oxide fuel cell stacks. Part II: Loss of gas-tightness, electrical contact and thermal buckling

Science.gov (United States)

Nakajo, Arata; Wuillemin, Zacharie; Van herle, Jan; Favrat, Daniel

Structural stability issues in planar solid oxide fuel cells arise from the mismatch between the coefficients of thermal expansion of the components. The stress state at operating temperature is the superposition of several contributions, which differ depending on the component. First, the cells accumulate residual stresses due to the sintering phase during the manufacturing process. Further, the load applied during assembly of the stack to ensure electric contact and flatten the cells prevents a completely stress-free expansion of each component during the heat-up. Finally, thermal gradients cause additional stresses in operation. The temperature profile generated by a thermo-electrochemical model implemented in an equation-oriented process modelling tool (gPROMS) was imported into finite-element software (ABAQUS) to calculate the distribution of stress and contact pressure on all components of a standard solid oxide fuel cell repeat unit. The different layers of the cell in exception of the cathode, i.e. anode, electrolyte and compensating layer were considered in the analysis to account for the cell curvature. Both steady-state and dynamic simulations were performed, with an emphasis on the cycling of the electrical load. The study includes two different types of cell, operation under both thermal partial oxidation and internal steam-methane reforming and two different initial thicknesses of the air and fuel compressive sealing gaskets. The results generated by the models are presented in two papers: Part I focuses on cell cracking. In the present paper, Part II, the occurrences of loss of gas-tightness in the compressive gaskets and/or electrical contact in the gas diffusion layer were identified. In addition, the dependence on temperature of both coefficients of thermal expansion and Young's modulus of the metallic interconnect (MIC) were implemented in the finite-element model to compute the plastic deformation, while the possibilities of thermal buckling

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

Science.gov (United States)

Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477

Energy Technology Data Exchange (ETDEWEB)

Hardin, Ernest [Sandia National Laboratories, P.O. Box 5800 MS 0736, Albuquerque, NM 87185 (United States); Blink, James [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551-0808 (United States); Carter, Joe [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States); Fratoni, Massimiliano; Greenberg, Harris; Sutton, Mark [Lawrence Livermore National Laboratory (United States); Howard, Robert [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2012-07-01

A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE), Used Fuel Disposition campaign. Reference concepts are identified for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. These were analyzed for waste inventory cases representing a range of waste types that could be produced by advanced nuclear fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress. All of these disposal concepts are enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. Enclosed modes have less capacity to dissipate heat than open modes such as that proposed for a repository at Yucca Mountain. Thermal analysis has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature limits for different disposal concepts. For the crystalline rock and clay/shale repository concepts, a waste package surface temperature limit of 100 deg. C was assumed to prevent changes in clay-based buffer material or clay-rich host rock. Surface decay storage of 50 to 100 years is needed for disposal of high-burnup LWR SNF in 4-PWR packages, or disposal of HLW glass from reprocessing LWR uranium oxide (UOX) fuel. High-level waste (HLW) from reprocessing of metal fuel used in a fast reactor could be disposed after decay storage of 50 years or less. For disposal in salt the rock thermal conductivity is significantly greater, and higher temperatures (200 deg. C) can be tolerated at the waste package surface. Decay storage of 10 years or less is needed for high-burnup LWR SNF in 4-PWR

Thermal performance of a meso-scale liquid-fuel combustor

International Nuclear Information System (INIS)

Vijayan, V.; Gupta, A.K.

2011-01-01

Research highlights: → Demonstrated successful combustion of liquid fuel-air mixtures in a novel meso-scale combustor. → Flame quenching was eliminated using heat recirculation in a swiss roll type combustor that also extended the flammability limits. → Liquid fuel was rapidly vaporized with the use of hot narrow channel walls that eliminated the need of a fuel atomizer. → Maximum power density of the combustor was estimated to be about 8.5 GW/m3 and heat load in the range of 50-280W. → Overall efficiency of the combustor was estimated in the range of 12 to 20%. - Abstract: Combustion in small scale devices poses significant challenges due to the quenching of reactions from wall heat losses as well as the significantly reduced time available for mixing and combustion. In the case of liquid fuels there are additional challenges related to atomization, vaporization and mixing with the oxidant in the very short time-scale liquid-fuel combustor. The liquid fuel employed here is methanol with air as the oxidizer. The combustor was designed based on the heat recirculating concept wherein the incoming reactants are preheated by the combustion products through heat exchange occurring via combustor walls. The combustor was fabricated from Zirconium phosphate, a ceramic with very low thermal conductivity (0.8 W m -1 K -1 ). The combustor had rectangular shaped double spiral geometry with combustion chamber in the center of the spiral formed by inlet and exhaust channels. Methanol and air were introduced immediately upstream at inlet of the combustor. The preheated walls of the inlet channel also act as a pre-vaporizer for liquid fuel which vaporizes the liquid fuel and then mixes with air prior to the fuel-air mixture reaching the combustion chamber. Rapid pre-vaporization of the liquid fuel by the hot narrow channel walls eliminated the necessity for a fuel atomizer. Self-sustained combustion of methanol-air was achieved in a chamber volume as small as 32.6 mm 3

Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Saxena, Aakanksha

2014-01-01

The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

Thermal conductivity of beginning-of-life uranium-plutonium mixed oxide fuel for fast reactor (Interim report)

International Nuclear Information System (INIS)

Inoue, Masaki; Mizuno, Tomoyasu; Asaga, Takeo

1997-11-01

Thermal conductivity of uranium-plutonium mixed oxide fuel for fast reactor at beginning-of-life was correlated based on the recent results in order to apply to the fuel design and the fuel performance analysis. A number of experimental results of unirradiated fuel specimens were corrected from open literatures and PNC internal reports and examined for the database. In this work two porosity correction factors were needed for high density fuel and low density fuel (around the current Monju specification). The universal porosity correction factor was not determined in this work. In the next step, theoretical and analytical considerations should be taken into account. (J.P.N.)

Thermally induced dispersion mechanisms for aluminum-based plate-type fuels under rapid transient energy deposition

International Nuclear Information System (INIS)

Georgevich, V.; Taleyarkham, R.P.; Navarro-Valenti, S.; Kim, S.H.

1995-01-01

A thermally induced dispersion model was developed to analyze for dispersive potential and determine onset of fuel plate dispersion for Al-based research and test reactor fuels. Effect of rapid energy deposition in a fuel plate was simulated. Several data types for Al-based fuels tested in the Nuclear Safety Research Reactor in Japan and in the Transient Reactor Test in Idaho were reviewed. Analyses of experiments show that onset of fuel dispersion is linked to a sharp rise in predicted strain rate, which futher coincides with onset of Al vaporization. Analysis also shows that Al oxidation and exothermal chemical reaction between the fuel and Al can significantly affect the energy deposition characteristics, and therefore dispersion onset connected with Al vaporization, and affect onset of vaporization

Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

International Nuclear Information System (INIS)

Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui

2016-01-01

Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

Energy Technology Data Exchange (ETDEWEB)

Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

2016-05-15

Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.