Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

Science.gov (United States)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

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.

The modeling and simulation of thermal based modified solid oxide fuel cell (SOFC for grid-connected systems

Directory of Open Access Journals (Sweden)

Ayetül Gelen

2015-05-01

Full Text Available This paper presents a thermal based modified dynamic model of a Solid Oxide Fuel Cell (SOFC for grid-connected systems. The proposed fuel cell model involves ohmic, activation and concentration voltage losses, thermal dynamics, methanol reformer, fuel utilization factor and power limiting module. A power conditioning unit (PCU, which consists of a DC-DC boost converter and a DC-AC voltage-source inverter (VSI, their controller, transformer and filter, is designed for grid-connected systems. The voltage-source inverter with six Insulated Gate Bipolar Transistor (IGBT switches inverts the DC voltage that comes from the converter into a sinusoidal voltage synchronized with the grid. The simulations and modeling of the system are developed on Matlab/Simulink environment. The performance of SOFC with converter is examined under step and random load conditions. The simulation results show that the designed boost converter for the proposed thermal based modified SOFC model has fairly followed different DC load variations. Finally, the AC bus of 400 Volt and 50 Hz is connected to a single-machine infinite bus (SMIB through a transmission line. The real and reactive power managements of the inverter are analyzed by an infinite bus system. Thus, the desired nominal values are properly obtained by means of the inverter controller.

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.

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)

Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

International Nuclear Information System (INIS)

Casal, V.; Malang, S.; Rust, K.

1982-10-01

For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

Thermal conductivity model of vibro-packed fuel

International Nuclear Information System (INIS)

Yeon Soo, Kim

2001-01-01

In an effort to dispose of excess weapons grade plutonium accumulated in the cold war era in the United States and the Russian Federation, one method currently under investigation is the conversion of the plutonium into mixed oxide (MOX) reactor fuel for LWRs and fast reactors in the Russian Federation. A fuel option already partly developed at the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad is that of vibro-packed MOX. Fuel rod fabrication using powder vibro-packing is attractive because it includes neither a process too complex to operate in glove boxes (or remotely), nor a waste-producing step necessary for the conventional pellet rod fabrication. However, because of its loose bonding between fuel particles at the beginning of life, vibro-packed MOX fuel has a somewhat less effective thermal conductivity than fully sintered pellet fuel, and undergoes more restructuring. Helium would also likely be pressurized in vibro-packed MOX fuel rods for LWRs to enhance initial fuel thermal conductivity. The combination of these two factors complicates development of an accurate thermal conductivity model. But clearly in order to predict fuel thermomechanical responses during irradiation of vibro-packed MOX fuel, fuel thermal conductivity must be known. The Vibropac fuel of interest in this study refers the fuel that is compacted with irregular fragments of mixed oxide fuel. In this paper, the thermal-conductivity models in the literature that dealt with relatively similar situations to the present case are examined. Then, the best model is selected based on accuracy of prediction and applicability. Then, the selected model is expanded to fit the various situations of interest. (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)

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

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

Science.gov (United States)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

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.

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.

Multiscale development of a fission gas thermal conductivity model: Coupling atomic, meso and continuum level simulations

International Nuclear Information System (INIS)

Tonks, Michael R.; Millett, Paul C.; Nerikar, Pankaj; Du, Shiyu; Andersson, David; Stanek, Christopher R.; Gaston, Derek; Andrs, David; Williamson, Richard

2013-01-01

Fission gas production and evolution significantly impact the fuel performance, causing swelling, a reduction in the thermal conductivity and fission gas release. However, typical empirical models of fuel properties treat each of these effects separately and uncoupled. Here, we couple a fission gas release model to a model of the impact of fission gas on the fuel thermal conductivity. To quantify the specific impact of grain boundary (GB) bubbles on the thermal conductivity, we use atomistic and mesoscale simulations. Atomistic molecular dynamic simulations were employed to determine the GB thermal resistance. These values were then used in mesoscale heat conduction simulations to develop a mechanistic expression for the effective GB thermal resistance of a GB containing gas bubbles, as a function of the percentage of the GB covered by fission gas. The coupled fission gas release and thermal conductivity model was implemented in Idaho National Laboratory’s BISON fuel performance code to model the behavior of a 10-pellet LWR fuel rodlet, showing how the fission gas impacts the UO 2 thermal conductivity. Furthermore, additional BISON simulations were conducted to demonstrate the impact of average grain size on both the fuel thermal conductivity and the fission gas release

Thermal behaviour of pressure tube under fully and partially voided heating conditions using 19 pin fuel element simulator

International Nuclear Information System (INIS)

Yadav, Ashwini K.; Kumar, Ravi; Gupta, Akhilesh; Chatterjee, B.; Mukhopadhya, D.; Lele, H.G.

2011-01-01

In a nuclear reactor temperature can rise drastically during LOCA due to failure of heat transportation system and subsequently leads to mechanical deformations like sagging, ballooning and breaching of pressure tube. To understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of 220 MWe Indian Pressurised Heavy Water Reactor (IPHWR). The symmetrical heating of pressure tube of 1 m length was done through resistance heating of 19 pins under 13.5 kW power using a rectifier and the variation of temperatures over the circumference of pressure tube (PT), calandria tube (CT) and clad tubes were measured. The sagging of pressure tube was initiated at 460 deg C temperature and highest temperature attained was 650 deg C. The highest temperature attained by clad tubes was 680 deg C (over outer ring) and heat is dissipated to calandria vessel mainly due to radiation and natural convection. Again to simulate partially voided conditions, asymmetrical heating of pressure was carried out by injecting 8 kW power to upper 8 pins of fuel simulator. A maximum temperature difference of 295 deg C was observed over the circumference of pressure tube which highlights the magnitude of thermal stresses and its role in breaching of pressure tube under partially voided conditions. Integrity of pressure tube was retained during both symmetrical and asymmetrical heatup conditions. (author)

Physical model of the nuclear fuel cycle simulation code SITON

International Nuclear Information System (INIS)

Brolly, Á.; Halász, M.; Szieberth, M.; Nagy, L.; Fehér, S.

2017-01-01

Finding answers to main challenges of nuclear energy, like resource utilisation or waste minimisation, calls for transient fuel cycle modelling. This motivation led to the development of SITON v2.0 a dynamic, discrete facilities/discrete materials and also discrete events fuel cycle simulation code. The physical model of the code includes the most important fuel cycle facilities. Facilities can be connected flexibly; their number is not limited. Material transfer between facilities is tracked by taking into account 52 nuclides. Composition of discharged fuel is determined using burnup tables except for the 2400 MW thermal power design of the Gas-Cooled Fast Reactor (GFR2400). For the GFR2400 the FITXS method is used, which fits one-group microscopic cross-sections as polynomial functions of the fuel composition. This method is accurate and fast enough to be used in fuel cycle simulations. Operation of the fuel cycle, i.e. material requests and transfers, is described by discrete events. In advance of the simulation reactors and plants formulate their requests as events; triggered requests are tracked. After that, the events are simulated, i.e. the requests are fulfilled and composition of the material flow between facilities is calculated. To demonstrate capabilities of SITON v2.0, a hypothetical transient fuel cycle is presented in which a 4-unit VVER-440 reactor park was replaced by one GFR2400 that recycled its own spent fuel. It is found that the GFR2400 can be started if the cooling time of its spent fuel is 2 years. However, if the cooling time is 5 years it needs an additional plutonium feed, which can be covered from the spent fuel of a Generation III light water reactor.

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)

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

Science.gov (United States)

Emrich, William J.; Moran, Robert P.; Pearson, J. Bose

2013-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities. Keywords: Nuclear Thermal Propulsion, Simulator

Thermal Conductivity Measurement and Analysis of Fully Ceramic Microencapsulated fuel

International Nuclear Information System (INIS)

Lee, H. G.; Kim, D. J.; Park, J. Y.; Kim, W. J.; Lee, S. J.

2015-01-01

FCM nuclear fuel is composed of tristructural isotropic(TRISO) fuel particle and SiC ceramic matrix. SiC ceramic matrix play an essential part in protecting fission product. In the FCM fuel concept, fission product is doubly protected by TRISO coating layer and SiC ceramic matrix in comparison with the current commercial UO2 fuel system of LWR. In addition to a safety enhancement of FCM fuel, thermal conductivity of SiC ceramic matrix is better than that of UO2 fuel. Because the centerline temperature of FCM fuel is lower than that of the current UO2 fuel due to the difference of thermal conductivity of fuel, an operational release of fission products from the fuel can be reduced. SiC ceramic has attracted for nuclear fuel application due to its high thermal conductivity properties with good radiation tolerant properties, a low neutron absorption cross-section and a high corrosion resistance. Thermal conductivity of ceramic matrix composite depends on the thermal conductivity of each component and the morphology of reinforcement materials such as fibers and particles. There are many results about thermal conductivity of fiber-reinforced composite like as SiCf/SiC composite. Thermal conductivity of SiC ceramics and FCM pellets with the volume fraction of TRISO particles were measured and analyzed by analytical models. Polycrystalline SiC ceramics and FCM pellets with TRISO particles were fabricated by hot press sintering with sintering additives. Thermal conductivity of the FCM pellets with TRISO particles of 0 vol.%, 10 vol.%, 20 vol.%, 30 vol.% and 40 vol.% show 68.4, 52.3, 46.8, 43.0 and 34.5 W/mK, respectively. As the volume fraction of TRISO particles increased, the measured thermal conductivity values closely followed the prediction of Maxwell's equation

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

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)

Modeling and Simulation of a Nuclear Fuel Element Test Section

Science.gov (United States)

Moran, Robert P.; Emrich, William

2011-01-01

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

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

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

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

International Nuclear Information System (INIS)

Kim, Seung Su; Ryu, Ho Jin

2015-01-01

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

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

Science.gov (United States)

Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

2012-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

Thermal conductivity of fresh and irradiated U-Mo fuels

Science.gov (United States)

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Elgeti, Stefan; Reiter, Christian; Robinson, Adam. B.; Smith, Frances. N.; Wachs, Daniel. M.; Petry, Winfried

2018-05-01

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.

DELOCA, a code for simulation of CANDU fuel channel in thermal transients

International Nuclear Information System (INIS)

Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.

2005-01-01

Full text: In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)

DELOCA, a code for simulation of CANDU fuel channel in thermal transients

International Nuclear Information System (INIS)

Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.

2005-01-01

In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)

Simulations of a PSD Plastic Neutron Collar for Assaying Fresh Fuel

Energy Technology Data Exchange (ETDEWEB)

Hausladen, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Newby, Jason [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-11-01

The potential performance of a notional active coincidence collar for assaying uranium fuel based on segmented detectors constructed from the new PSD plastic fast organic scintillator with pulse shape discrimination capability was investigated in simulation. Like the International Atomic Energy Agency's present Uranium Neutron Collar for LEU (UNCL), the PSD plastic collar would also function by stimulating fission in the ²³⁵U content of the fuel with a moderated ²⁴¹Am/Li neutron source and detecting instances of induced fission via neutron coincidence counting. In contrast to the moderated detectors of the UNCL, the fast time scale of detection in the scintillator eliminates statistical errors due to accidental coincidences that limit the performance of the UNCL. However, the potential to detect a single neutron multiple times historically has been one of the properties of organic scintillator detectors that has prevented their adoption for international safeguards applications. Consequently, as part of the analysis of simulated data, a method was developed by which true neutron-neutron coincidences can be distinguished from inter-detector scatter that takes advantage of the position and timing resolution of segmented detectors. Then, the performance of the notional simulated coincidence collar was evaluated for assaying a variety of fresh fuels, including some containing burnable poisons and partial defects. In these simulations, particular attention was paid to the analysis of fast mode measurements. In fast mode, a Cd liner is placed inside the collar to shield the fuel from the interrogating source and detector moderators, thereby eliminating the thermal neutron flux that is most sensitive to the presence of burnable poisons that are ubiquitous in modern nuclear fuels. The simulations indicate that the predicted precision of fast mode measurements is similar to what can be achieved by the present UNCL in thermal mode. For example, the

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

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

Advanced thermally stable jet fuels

Energy Technology Data Exchange (ETDEWEB)

Schobert, H.H.

1999-01-31

The Pennsylvania State University 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; (5) Assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Future high-Mach aircraft will place severe thermal demands on jet fuels, requiring the development of novel, hybrid fuel mixtures capable of withstanding temperatures in the range of 400--500 C. In the new aircraft, jet fuel will serve as both an energy source and a heat sink for cooling the airframe, engine, and system components. The ultimate development of such advanced fuels requires a thorough understanding of the thermal decomposition behavior of jet fuels under supercritical conditions. Considering that jet fuels consist of hundreds of compounds, this task must begin with a study of the thermal degradation behavior of select model compounds under supercritical conditions. The research performed by The Pennsylvania State University was focused on five major tasks that reflect the objectives stated above: Task 1: Investigation of the Quantitative Degradation of Fuels; Task 2: Investigation of Incipient Deposition; Task 3: Characterization of Solid Gums, Sediments, and Carbonaceous Deposits; Task 4: Coal-Based Fuel Stabilization Studies; and Task 5: Exploratory Studies on the Direct Conversion of Coal to High Quality Jet Fuels. The major findings of each of these tasks are presented in this executive summary. A description of the sub-tasks performed under each of these tasks and the findings of those studies are provided in the remainder of this volume

Simulation of thermal phenomena expected in fuel coolant interactions in LMFBRs

International Nuclear Information System (INIS)

Yasin, J.

1976-12-01

High pressures and mechanical work may result when thermal energy is transferred from molten fuel to the coolant in a Liquid Metal Fast Breeder Reactor core meltdown accident. Two aspects of the interaction are examined in the thesis. First, the formation of high pressure pulses termed ''Vapor Explosions,'' and second, the distribution of the molten material into smaller particles, termed ''Fragmentation'', are studied. To understand the nature of the interaction simulant materials were used. Molten bismuth, molten tin and molten glass were dropped into water under various conditions. The interactions were recorded using multiflash and high speed photographing techniques. The pressure pulses were measured using transducers and the debris was examined by photographing them with an electron microscope. It was observed that vapor explosions have thresholds which depend on the material being dropped, its temperature and the bath conditions. The vapor explosions were enhanced by stratifying the bath. It was also noticed that the intensity of the vapor explosion depends on the way the molten drop fragmented in the initial stages of the interaction. The experiments with glass showed that the mode of fragmentation is important in determining when and if a vapor explosion is to be expected. The glass fragmented extensively but without any accompanying vapor explosion. The electron microscope photographs of the glass debris showed that thermal stress and surface tension phenomenon are apparently the cause of the fragmentation

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

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)

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

discontinuities in both temperature and displacement fields at crack locations has been developed and is being applied to thermal fracture of LWR fuel. A DEM model of coupled heat conduction and solid mechanics has been developed and used to simulate random initiation and propagation of thermally driven cracks during initial power cycles. This DEM model predicts the formation of realistic radial cracking patterns during power rise and circumferential cracks as power is ramped down. These initial results are very encouraging, and these techniques are expected to provide improved understanding of fuel behavior in a wide variety of conditions.

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 properties of heterogeneous fuels

International Nuclear Information System (INIS)

Staicu, D.; Beauvy, M.

1998-01-01

Fresh or irradiated nuclear fuels are composites or solid solutions more or less heterogeneous, and their thermal conductivities are strongly dependent on the microstructure. The effective thermal conductivities of these heterogeneous solids must be determined for the modelling of the behaviour under irradiation. Different methods (analytical or numerical) published in the literature can be used for the calculation of this effective thermal conductivity. They are analysed and discussed, but finally only few of them are really useful because the assumptions selected are often not compatible with the complex microstructures observed in the fuels. Numerical calculations of the effective thermal conductivity of various fuels based on the microstructure information provided in our laboratory by optical microscopy or electron micro-probe analysis images, have been done for the validation of these methods. The conditions necessary for accurate results on effective thermal conductivity through these numerical calculations are discussed. (author)

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.)

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

Large Eddy Simulation of turbulent flow in wire wrapped fuel pin bundles cooled by sodium

International Nuclear Information System (INIS)

Saxena, Aakanksha; Cadiou, Thierry; Bieder, Ulrich; Viazzo, Stephane

2013-06-01

The objective of the study is to understand the thermal hydraulics in a core sub-assembly with liquid sodium as coolant by performing detailed numerical simulations. The passage for the coolant flow between the fuel rods is maintained by thin wires wrapped around the rods. The contact point between the fuel pin and the spacer wire is the region of creation of hot spots and a cyclic variation of temperature in hot spots can adversely affect the mechanical properties of the clad due to the phenomena like thermal stripping. The current status quo provides two different models to perform the numerical simulations, namely Reynolds Averaged Navier-Stokes (RANS) and Large Eddy Simulation (LES). The two models differ in the extent of modelling used to close the Navier-Stokes equations. LES is a filtered approach where the large scale of motions are explicitly resolved while the small scale motions are modelled whereas RANS is a time averaging approach where all scale of motions are modelled. Thus LES involves less modelling as compared to RANS and so the results are comparatively more accurate. An attempt has been made to use the LES model. The simulations have been performed using the code Trio-U (developed by CEA). The turbulent statistics of the flow and thermal quantities are calculated. Finally the goal is to obtain the frequency of temperature oscillations at the region of hot spots near the spacer wire. (authors)

Effect of point defects on the thermal conductivity of UO2: molecular dynamics simulations

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiang-Yang [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)

2015-07-21

The thermal conductivity of uranium dioxide (UO₂) fuel is an important materials property that affects fuel performance since it is a key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. [1] The thermal conductivity of UO₂ nuclear fuel is also affected by fission gas, fission products, defects, and microstructural features such as grain boundaries. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of irradiation induced point defects on the thermal conductivity of UO₂, as a function of defect 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 [2].

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 simulation of the magnesium thermal of metallic uranium reduction

International Nuclear Information System (INIS)

Borges, W.A.; Saliba-Silva, A.M.

2008-01-01

Metallic uranium production is vital to fabricate fuel elements for nuclear research reactors and to produce radioisotopes and radiopharmaceuticals. Metallic uranium is got via magnesiothermal reduction of UF 4 . This reaction is carried out inside a closed graphite crucible inserted in a metallic reactor adequately sealed without any outside contact. The assembled set is gradually heated up inside a pit furnace up to reach the reaction ignition temperature (between 600-650 deg C). The optimization of the reactive system depends on the mathematical modeling using simulation by finite elements and computational calculation with specialized programs. In this way, the reactants' thermal behavior is forecast until they reach the ignition temperature. The optimization of the uranium production reaction is based on minimization of thermal losses using better the exo thermal reaction heat. As lower the thermal losses, as higher would be the heat amount to raise the temperature of reaction products. This promotes the adequate melting of uranium and slag, so allowing better metal/slag separation with higher metallic yield. This work shows how the mathematical simulation is made and supplies some preliminary results. (author)

Phase-field simulations of pore migration and morphology change in thermal gradients

Energy Technology Data Exchange (ETDEWEB)

Vance, Ian W.; Millett, Paul C., E-mail: pmillett@uark.edu

2017-07-15

Here we present a phase-field simulation model that captures the thermal-gradient-driven migration of pores in oxide fuel associated with fuel restructuring. The model utilizes a Cahn-Hilliard equation supplemented with an advection term to describe the vapor transport of fuel material through the pore interior due to gradients in vapor pressure. Simulations demonstrate that the model not only predicts pore migration towards the centerline of the fuel, but also a concurrent change in pore shape during migration from an initially isotropic morphology to either a lenticular morphology or a prolate morphology depending on the vapor transport conditions. This model is a necessary first step to conducting accurate simulations of the microscopic changes that occur during the complicated process of oxide fuel restructuring.

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

Simulation of a 250 kW diesel fuel processor/PEM fuel cell system

Science.gov (United States)

Amphlett, J. C.; Mann, R. F.; Peppley, B. A.; Roberge, P. R.; Rodrigues, A.; Salvador, J. P.

Polymer-electrolyte membrane (PEM) fuel cell systems offer a potential power source for utility and mobile applications. Practical fuel cell systems use fuel processors for the production of hydrogen-rich gas. Liquid fuels, such as diesel or other related fuels, are attractive options as feeds to a fuel processor. The generation of hydrogen gas for fuel cells, in most cases, becomes the crucial design issue with respect to weight and volume in these applications. Furthermore, these systems will require a gas clean-up system to insure that the fuel quality meets the demands of the cell anode. The endothermic nature of the reformer will have a significant affect on the overall system efficiency. The gas clean-up system may also significantly effect the overall heat balance. To optimize the performance of this integrated system, therefore, waste heat must be used effectively. Previously, we have concentrated on catalytic methanol-steam reforming. A model of a methanol steam reformer has been previously developed and has been used as the basis for a new, higher temperature model for liquid hydrocarbon fuels. Similarly, our fuel cell evaluation program previously led to the development of a steady-state electrochemical fuel cell model (SSEM). The hydrocarbon fuel processor model and the SSEM have now been incorporated in the development of a process simulation of a 250 kW diesel-fueled reformer/fuel cell system using a process simulator. The performance of this system has been investigated for a variety of operating conditions and a preliminary assessment of thermal integration issues has been carried out. This study demonstrates the application of a process simulation model as a design analysis tool for the development of a 250 kW fuel cell system.

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.

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

International Nuclear Information System (INIS)

Yaksh, M.; Wang, C.

2004-01-01

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

ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

International Nuclear Information System (INIS)

1990-10-01

The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

Large-scale numerical simulations on two-phase flow behavior in a fuel bundle of RMWR with the earth simulator

International Nuclear Information System (INIS)

Kazuyuki, Takase; Hiroyuki, Yoshida; Hidesada, Tamai; Hajime, Akimoto; Yasuo, Ose

2003-01-01

Fluid flow characteristics in a fuel bundle of a reduced-moderation light water reactor (RMWR) with a tight-lattice core were analyzed numerically using a newly developed two-phase flow analysis code under the full bundle size condition. Conventional analysis methods such as sub-channel codes need composition equations based on the experimental data. In case that there are no experimental data regarding to the thermal-hydraulics in the tight-lattice core, therefore, it is difficult to obtain high prediction accuracy on the thermal design of the RMWR. Then the direct numerical simulations with the earth simulator were chosen. The axial velocity distribution in a fuel bundle changed sharply around a grid spacer and its quantitative evaluation was obtained from the present preliminary numerical study. The high prospect was acquired on the possibility of establishment of the thermal design procedure of the RMWR by large-scale direct simulations. (authors)

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

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.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

International Nuclear Information System (INIS)

Emrich, William J. Jr.

2008-01-01

To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts

Industrial thermoforming simulation of automotive fuel tanks

International Nuclear Information System (INIS)

Wiesche, Stefan aus der

2004-01-01

An industrial thermoforming simulation with regard to automotive plastic fuel tanks is presented including all relevant process stages. The radiative and conductive heat transfer during the reheat stage, the deformation and stress behaviour during the forming stage, and the final cooling stage are simulated. The modelling of the thermal and rheological behaviour of the involved material is investigated in greater detail. By means of experimental data it is found that modelling of the phase transition during the process is highly important for predicting correct wall thickness distributions

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.

DNA - A Thermal Energy System Simulator

DEFF Research Database (Denmark)

2008-01-01

DNA is a general energy system simulator for both steady-state and dynamic simulation. The program includes a * component model library * thermodynamic state models for fluids and solid fuels and * standard numerical solvers for differential and algebraic equation systems and is free and portable...... (open source, open use, standard FORTRAN77). DNA is text-based using whichever editor, you like best. It has been integerated with the emacs editor. This is usually available on unix-like systems. for windows we recommend the Installation instructions for windows: First install emacs and then run...... the DNA installer...

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

Simulations of beam-fueled supershot-like plasmas near ignition

International Nuclear Information System (INIS)

Budny, R.V.; Grisham, L.; Jassby, D.L.

1992-01-01

In certain conditions, neutral beam injection (NBI) and low recycling result in supershot plasmas. These are characterized by peaked density profiles and high central ion temperatures. We discuss the potential advantages of NBI fueled supershot-like plasmas in tokamaks operating near ignition. The goal is to investigate the feasibility of these plasmas to aid in the design of future advanced tokamaks. NBI has been very successful in advancing tokamak plasmas close to ignition conditions. The primary benefits of NBI are heating and particle fueling, but the plasma currents generated by the beam ions are also of considerable interest. The optimal injection energy E inj for the beam ions depends on the desired role of the NBI. For central particle fueling, E inj should be low to maximize the particle current at fixed P B , but high enough to penetrate to the center. For heating and current drive, higher E inj is preferable for deepest penetration. With the standard positive ion beam technology, the neutralization efficiency becomes too low for useful power densities if E inj is significantly greater than about 120 keV. Negative ion beam sources would be useful for heating and current drive at very high E inj (500 keV or more), but the fueling rate of NBI is too low to be practical. It seems generally accepted that future tokamaks which operate closer to ignition will have to be fueled and heated by means other than NBI since it is argued that the beams with low E inj cannot penetrate deeply into the dense plasmas of interest. (author) 3 refs., 4 figs

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.

Simulation of the injection casting of metallic fuels

International Nuclear Information System (INIS)

Nakagawa, Tomokazu; Ogata, Takanari; Tokiwai, Moriyasu.

1989-01-01

For the fabrication of metallic fuel pins, injection casting is a preferable process because the simplicity of the process is suitable for remote operation. In this process, the molten metal in the crucible is injected into evacuated molds (suspended above the crucible) by pressurizing the casting furnace. Argonne National Laboratory has already adopted this process in the Integral Fast Reactor program. To obtain fuel pins with good quality, the casting parameters, such as the molten metal temperature, the magnitude of the pressure applied, the pressurizing rate, the cooling time, etc., must be optimized. Otherwise, bad-quality castings (short castings, rough surfaces, shrinkage cavities, mold fracture) may result. Therefore, it is very important in designing the casting equipment and optimizing the operation conditions to be able to predict the fluid and thermal behavior of the castings. This paper describes methods to simulate the heat and mass transfer in the molds and molten metallic fuel during injection casting. The results obtained by simulation are compared with experimental ones. Also, appropriate casting conditions for the uranium-plutonium-zirconium alloy are discussed based on the simulated results

Experiments and simulations of NOx formation in the combustion of hydroxylated fuels

KAUST Repository

Bohon, Myles

2015-06-01

This work investigates the influence of molecular structure in hydroxylated fuels (i.e. fuels with one or more hydroxyl groups), such as alcohols and polyols, on NOx formation. The fuels studied are three lower alcohols (methanol, ethanol, and n-propanol), two diols (1,2-ethanediol and 1,2-propanediol), and one triol (1,2,3-propanetriol); all of which are liquids at room temperature and span a wide range of thermophysical properties. Experimental stack emissions measurements of NO/NO2, CO, and CO2 and flame temperature profiles utilizing a rake of thermocouples were obtained in globally lean, swirling, liquid atomized spray flames inside a refractory-lined combustion chamber as a function of the atomizing air flow rate and swirl number. These experiments show significantly lower NOx formation with increasing fuel oxygen content despite similarities in the flame temperature profiles. By controlling the temperature profiles, the contribution to NOx formation through the thermal mechanism were matched, and variations in the contribution through non-thermal NOx formation pathways are observed. Simulations in a perfectly stirred reactor, at conditions representative of those measured within the combustion region, were conducted as a function of temperature and equivalence ratio. The simulations employed a detailed high temperature chemical kinetic model for NOx formation from hydroxylated fuels developed based on recent alcohol combustion models and extended to include polyol combustion chemistry. These simulations provide a qualitative comparison to the range of temperatures and equivalence ratios observed in complex swirling flows and provide insight into the influence of variations in the fuel decomposition pathways on NOx formation. It is observed that increasing the fuel bound oxygen concentration ultimately reduces the formation of NOx by increasing the proportion of fuel oxidized through formaldehyde, as opposed to acetylene or acetaldehyde. The subsequent

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

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

Solar thermal power plants simulation using the TRNSYS software

Energy Technology Data Exchange (ETDEWEB)

Popel, O.S.; Frid, S.E.; Shpilrain, E.E. [Institute for High Temperatures, Russian Academy of Sciences (IVTAN), Moscow (Russian Federation)

1999-03-01

The paper describes activity directed on the TRNSYS software application for mathematical simulation of solar thermal power plants. First stage of developments has been devoted to simulation and thermodynamic analysis of the Hybrid Solar-Fuel Thermal Power Plants (HSFTPP) with gas turbine installations. Three schemes of HSFTPP, namely: Gas Turbine Regenerative Cycle, Brayton Cycle with Steam Injection and Combined Brayton-Rankine Cycle,- have been assembled and tested under the TRNSYS. For this purpose 18 new models of the schemes components (gas and steam turbines, compressor, heat-exchangers, steam generator, solar receiver, condenser, controllers, etc) have been elaborated and incorporated into the TRNSYS library of 'standard' components. The authors do expect that this initiative and received results will stimulate experts involved in the mathematical simulation of solar thermal power plants to join the described activity to contribute to acceleration of development and expansion of 'Solar Thermal Power Plants' branch of the TRNSYS. The proposed approach could provide an appropriate basis for standardization of analysis, models and assumptions for well-founded comparison of different schemes of advanced solar power plants. (authors)

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

Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

International Nuclear Information System (INIS)

Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki

2004-03-01

The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

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)

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)

SEFLEX - fuel rod simulator effects in flooding experiments. Pt. 2

International Nuclear Information System (INIS)

Ihle, P.; Rust, K.

1986-03-01

This report presents typical data and a limited heat transfer analysis from unblocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5 x 5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5 x 5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

International Nuclear Information System (INIS)

Ihle, P.; Rust, K.

1986-03-01

This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

Results from Tight and Loose Coupled Multiphysics in Nuclear Fuels Performance Simulations using BISON

Energy Technology Data Exchange (ETDEWEB)

S. R. Novascone; B. W. Spencer; D. Andrs; R. L. Williamson; J. D. Hales; D. M. Perez

2013-05-01

The behavior of nuclear fuel in the reactor environment is affected by multiple physics, most notably heat conduction and solid mechanics, which can have a strong influence on each other. To provide credible solutions, a fuel performance simulation code must have the ability to obtain solutions for each of the physics, including coupling between them. Solution strategies for solving systems of coupled equations can be categorized as loosely-coupled, where the individual physics are solved separately, keeping the solutions for the other physics fixed at each iteration, or tightly coupled, where the nonlinear solver simultaneously drives down the residual for each physics, taking into account the coupling between the physics in each nonlinear iteration. In this paper, we compare the performance of loosely and tightly coupled solution algorithms for thermomechanical problems involving coupled thermal and mechanical contact, which is a primary source of interdependence between thermal and mechanical solutions in fuel performance models. The results indicate that loosely-coupled simulations require significantly more nonlinear iterations, and may lead to convergence trouble when the thermal conductivity of the gap is too small. We also apply the tightly coupled solution strategy to a nuclear fuel simulation of an experiment in a test reactor. Studying the results from these simulations indicates that perhaps convergence for either approach may be problem dependent, i.e., there may be problems for which a loose coupled approach converges, where tightly coupled won’t converge and vice versa.

Results from tight and loose coupled multiphysics in nuclear fuels performance simulations using BISON

International Nuclear Information System (INIS)

Novascone, S. R.; Spencer, B. W.; Andrs, D.; Williamson, R. L.; Hales, J. D.; Perez, D. M.

2013-01-01

The behavior of nuclear fuel in the reactor environment is affected by multiple physics, most notably heat conduction and solid mechanics, which can have a strong influence on each other. To provide credible solutions, a fuel performance simulation code must have the ability to obtain solutions for each of the physics, including coupling between them. Solution strategies for solving systems of coupled equations can be categorized as loosely-coupled, where the individual physics are solved separately, keeping the solutions for the other physics fixed at each iteration, or tightly coupled, where the nonlinear solver simultaneously drives down the residual for each physics, taking into account the coupling between the physics in each nonlinear iteration. In this paper, we compare the performance of loosely and tightly coupled solution algorithms for thermomechanical problems involving coupled thermal and mechanical contact, which is a primary source of interdependence between thermal and mechanical solutions in fuel performance models. The results indicate that loosely-coupled simulations require significantly more nonlinear iterations, and may lead to convergence trouble when the thermal conductivity of the gap is too small. We also apply the tightly coupled solution strategy to a nuclear fuel simulation of an experiment in a test reactor. Studying the results from these simulations indicates that perhaps convergence for either approach may be problem dependent, i.e., there may be problems for which a loose coupled approach converges, where tightly coupled won't converge and vice versa. (authors)

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.

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

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

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)

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.

Characterizing high-temperature deformation of internally heated nuclear fuel element simulators

Energy Technology Data Exchange (ETDEWEB)

Belov, A.I.; Fong, R.W.L.; Leitch, B.W.; Nitheanandan, T.; Williams, A., E-mail: alexander.belov@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

The sag behaviour of a simulated nuclear fuel element during high-temperature transients has been investigated in an experiment utilizing an internal indirect heating method. The major motivation of the experiment was to improve understanding of the dominant mechanisms underlying the element thermo-mechanical response under loss-of-coolant accident conditions and to obtain accurate experimental data to support development of 3-D computational fuel element models. The experiment was conducted using an electrically heated CANDU fuel element simulator. Three consecutive thermal cycles with peak temperatures up to ≈1000 {sup o}C were applied to the element. The element sag deflections and sheath temperatures were measured. On heating up to 600 {sup o}C, only minor lateral deflections of the element were observed. Further heating to above 700 {sup o}C resulted in an element multi-rate creep and significant permanent bow. Post-test visual and X-ray examinations revealed a pronounced necking of the sheath at the pellet-to-pellet interface locations. A wall thickness reduction was detected in the necked region that is interpreted as a sheath longitudinal strain localization effect. The sheath cross-sectioning showed signs of a 'hard' pellet-cladding interaction due to the applied cycles. A 3-D model of the experiment was generated using the ANSYS finite element code. As a fully coupled thermal mechanical simulation is computationally expensive, it was deemed sufficient to use the measured sheath temperatures as a boundary condition, and thus an uncoupled mechanical simulation only was conducted. The ANSYS simulation results match the experiment sag observations well up to the point at which the fuel element started cooling down. (author)

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...

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

Full Text Available It is necessary to carry out fuel temperature preparation for space launch vehicles using hydrocarbon propellant components. A required temperature is reached with cooling or heating hydrocarbon fuel in ground facilities fuel storages. Fuel temperature preparing processes are among the most energy-intensive and lengthy processes that require the optimal technologies and regimes of cooling (heating fuel, which can be defined using the simulation of heat exchange processes for preparing the rocket fuel.The issues of research of different technologies and simulation of cooling processes of rocket fuel with liquid nitrogen are given in [1-10]. Diagrams of temperature preparation of hydrocarbon fuel, mathematical models and characteristics of cooling fuel with its direct contact with liquid nitrogen dispersed are considered, using the numerical solution of a system of heat transfer equations, in publications [3,9].Analytical models, allowing to determine the necessary flow rate and the mass of liquid nitrogen and the cooling (heating time fuel in specific conditions and requirements, are preferred for determining design and operational characteristics of the hydrocarbon fuel cooling system.A mathematical model of the temperature preparation processes is developed. Considered characteristics of these processes are based on the analytical solutions of the equations of heat transfer and allow to define operating parameters of temperature preparation of hydrocarbon fuel in the design and operation of the filling system of launch vehicles.The paper considers a technological system to fill the launch vehicles providing the temperature preparation of hydrocarbon gases at the launch site. In this system cooling the fuel in the storage tank before filling the launch vehicle is provided by hydrocarbon fuel bubbling with liquid nitrogen. Hydrocarbon fuel is heated with a pumping station, which provides fuel circulation through the heat exchanger-heater, with

Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

International Nuclear Information System (INIS)

Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

2005-01-01

Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

Science.gov (United States)

Emrich, William J., Jr.; Pearson, J. Boise; Schoenfeld, Michael P.

2015-01-01

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at

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.

COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM

Directory of Open Access Journals (Sweden)

Filip Osusky

2018-05-01

Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.

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

Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

Energy Technology Data Exchange (ETDEWEB)

Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

2017-02-15

Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

A MULTIDIMENSIONAL AND MULTIPHYSICS APPROACH TO NUCLEAR FUEL BEHAVIOR SIMULATION

Energy Technology Data Exchange (ETDEWEB)

R. L. Williamson; J. D. Hales; S. R. Novascone; M. R. Tonks; D. R. Gaston; C. J. Permann; D. Andrs; R. C. Martineau

2012-04-01

Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non-axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our the multidimensional, multiphysics approach to analyze a missing pellet surface problem. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

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

High Fidelity BWR Fuel Simulations

Energy Technology Data Exchange (ETDEWEB)

Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

Simulation of hydrogen releases from fuel-cell vehicles in tunnels

Energy Technology Data Exchange (ETDEWEB)

Houf, William G.; Evans, Greg H.; James, Scott C. [Sandia National Labs., Livermore, CA (United States); Merilo, Erik; Groethe, Mark [SRI International, Menlo Park, CA (United States)

2010-07-01

Simulation results for a hydrogen fuel-cell vehicle in a full-scale tunnel have been performed for the case where hydrogen gas is vented from the vehicle as a result of thermal activation of the pressure relief device (PRD). The same modeling approach used in the full-scale tunnel modeling was validated in a scaled model by comparing simulated results with measured results from a series of scaled-tunnel test experiments performed at the SRI Corral Hollow test facility. Results of the simulations were found to be in good agreement with the experimental data. Finally, a rudimentary risk analysis indicated that the level of potential risk from hydrogen vehicles accidents involving thermally activated PRDs in tunnels does not appear to significantly increase the current level of individual risk to the public from everyday life. (orig.)

Heaters to simulate fuel pins for heat transfer tests in single-phase liquid-metal-flow

International Nuclear Information System (INIS)

Casal, V.; Graf, E.; Hartmann, W.

1976-09-01

The development of heaters for thermal simulation of the fuel elements of liquid metal cooled fast breeder reactors (SNR) is reported. Beginning with the experimental demands various heating methods are discussed for thermodynamic investigations of the heat transfer in liquid metals. Then a preferred heater rod is derived to simulate the fuel pins of a SNR. Finally it is reported on the fabrication and the operation practice. (orig.) [de

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)

Metallic Fuel Casting Development and Parameter Optimization Simulations

International Nuclear Information System (INIS)

Fielding, Randall S.; Kennedy, J.R.; Crapps, J.; Unal, C.

2013-01-01

Conclusions: • Gravity casting is a feasible process for casting of metallic fuels: – May not be as robust as CGIC, more parameter dependent to find right “sweet spot” for high quality castings; – Fluid flow is very important and is affected by mold design, vent size, super heat, etc.; – Pressure differential assist was found to be detrimental. • Simulation found that vent location was important to allow adequate filling of mold; • Surface tension plays an important role in determining casting quality; • Casting and simulations high light the need for better characterized fluid physical and thermal properties; • Results from simulations will be incorporated in GACS design such as vent location and physical property characterization

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

Gelled fuel simulant

International Nuclear Information System (INIS)

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

1980-01-01

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

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.

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)

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.

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

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

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.

Nuclear Thermal Rocket Simulation in NPSS

Science.gov (United States)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

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)

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.

Casting thermal simulation

International Nuclear Information System (INIS)

Shamsuddin bin Sulaiman

1994-01-01

The whole of this study is concerned with process simulation in casting processes. This study describes the application of the finite element method as an aid to simulating the thermal design of a high pressure die casting die by analysing the cooling transients in the casting cycle. Two types of investigation were carried out to model the linear and non-linear cooling behavior with consideration of a thermal interface effect. The simulated cooling for different stages were presented in temperature contour form. These illustrate the successful application of the Finite Element Method to model the process and they illustrate the significance of the thermal interface at low pressure

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.

The simulation of CANDU fuel channel behavior in thermal transient conditions

International Nuclear Information System (INIS)

Mihalache, M.; Roth, M.; Radu, V.; Dumitrescu, I.

2005-01-01

In certain LOCA conditions into the CANDU fuel channel, is possible the ballooning of the pressure tube and the contact with the calandria tube. After the contact moment, a radial heat transfer to the moderator through the contact area is occurs. When the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. Thus, the fuel channel could lose its integrity. This paper present a computer code, DELOCA, developed in INR, which simulate the transient thermo-mechanical behaviour of CANDU fuel channel before and after contact. The code contains few models: alloy creep, heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. It was verified step by step by Contact1 and Cathena codes. In this paper, the results obtained at different temperature increasing rates are presented. Also, the contact moment for a RIH 5% postulated accident was presented. The input data was furnished by the Cathena thermo-hydraulic code. (author)

Thermal Analysis of Surrogate Simulated Molten Salts with Metal Chloride Impurities for Electrorefining Used Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Toni Y. Gutknecht; Guy L. Fredrickson; Vivek Utgikar

2012-04-01

This project is a fundamental study to measure thermal properties (liquidus, solidus, phase transformation, and enthalpy) of molten salt systems of interest to electrorefining operations, which are used in both the fuel cycle research & development mission and the spent fuel treatment mission of the Department of Energy. During electrorefining operations the electrolyte accumulates elements more active than uranium (transuranics, fission products and bond sodium). The accumulation needs to be closely monitored because the thermal properties of the electrolyte will change as the concentration of the impurities increases. During electrorefining (processing techniques used at the Idaho National Laboratory to separate uranium from spent nuclear fuel) it is important for the electrolyte to remain in a homogeneous liquid phase for operational safeguard and criticality reasons. The phase stability of molten salts in an electrorefiner may be adversely affected by the buildup of fission products in the electrolyte. Potential situations that need to be avoided are: (i) build up of fissile elements in the salt approaching the criticality limits specified for the vessel (ii) freezing of the salts due to change in the liquidus temperature and (iii) phase separation (non-homogenous solution) of elements. The stability (and homogeneity) of the phases can potentially be monitored through the thermal characterization of the salts, which can be a function of impurity concentration. This work describes the experimental results of typical salts compositions, consisting of chlorides of strontium, samarium, praseodymium, lanthanum, barium, cerium, cesium, neodymium, sodium and gadolinium (as a surrogate for both uranium and plutonium), used in the processing of used nuclear fuels. Differential scanning calorimetry was used to analyze numerous salt samples providing results on the thermal properties. The property of most interest to pyroprocessing is the liquidus temperature. It was

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

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

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

Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M.

2013-01-01

In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M., E-mail: amir@cdtn.br, E-mail: hcr@cdtn.br, E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br, E-mail: campolina@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/UFMG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel Artur P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2013-07-01

In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

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.

Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

Energy Technology Data Exchange (ETDEWEB)

Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)

2017-04-15

The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

International Nuclear Information System (INIS)

Williamson, R.L.

2011-01-01

Highlights: → The ABAQUS thermomechanics code is enhanced to enable simulation of nuclear fuel behavior. → Comparisons are made between discrete and smeared fuel pellet analysis. → Multidimensional and multipellet analysis is important for accurate prediction of PCMI. → Fully coupled thermomechanics results in very smooth prediction of fuel-clad gap closure. → A smeared-pellet approximation results in significant underprediction of clad radial displacements and plastic strain. - Abstract: A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO 2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

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.

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

Thermal hydraulic-Mechanic Integrated Simulation for Advanced Cladding Thermal Shock Fracture Analysis during Reflood Phase in LBLOCA

Energy Technology Data Exchange (ETDEWEB)

Son, Seong Min; Lee, You Ho; Cho, Jae Wan; Lee, Jeong Ik [KAERI, Daejeon (Korea, Republic of)

2016-05-15

This study suggested thermal hydraulic-mechanical integrated stress based methodology for analyzing the behavior of ATF type claddings by SiC-Duplex cladding LBLOCA simulation. Also, this paper showed that this methodology could predict real experimental result well. That concept for enhanced safety of LWR called Advanced Accident-Tolerance Fuel Cladding (ATF cladding, ATF) is researched actively. However, current nuclear fuel cladding design criteria for zircaloy cannot be apply to ATF directly because those criteria are mainly based on limiting their oxidation. So, the new methodology for ATF design criteria is necessary. In this study, stress based analysis methodology for ATF cladding design criteria is suggested. By simulating LBLOCA scenario of SiC cladding which is the one of the most promising candidate of ATF. Also we'll confirm our result briefly through comparing some facts from other experiments. This result is validating now. Some of results show good performance with 1-D failure analysis code for SiC fuel cladding that already developed and validated by Lee et al,. It will present in meeting. Furthermore, this simulation presented the possibility of understanding the behavior of cladding deeper. If designer can predict the dangerous region and the time precisely, it may be helpful for designing nuclear fuel cladding geometry and set safety criteria.

Evaluation of the effective thermal conductivity of UO{sub 2} fuel by combining Potts model and finite difference method

Energy Technology Data Exchange (ETDEWEB)

Oh, Jae-Yong, E-mail: tylor@kaeri.re.kr [Korea Atomic Energy Research Institute, Daedeok-daero 1045, Yuseong, Daejeon 305-353 (Korea, Republic of); Koo, Yang-Hyun; Lee, Byung-Ho; Tahk, Young-Wook [Korea Atomic Energy Research Institute, Daedeok-daero 1045, Yuseong, Daejeon 305-353 (Korea, Republic of)

2011-07-15

This paper evaluated the effects of porosity on the effective thermal conductivity of UO{sub 2} fuel by combining the Potts model and the finite difference method (FDM). Two types of microstructures representing irradiated UO{sub 2} microstructures were simulated by the Potts model in the three dimensional cubic system. One represented very small intragranular bubbles and a few intergranular bubbles under a low temperature condition. The other represented large intergranular bubbles under a high temperature or annealing condition. For the simulated microstructures, the effective thermal conductivities were determined by FDM calculation of the temperature distributions under steady state condition. They were compared with an experimental equation and the effect of bubble morphology was investigated by fitting a porosity shape factor in the Maxwell-Eucken equation. The simulation results showed a good agreement with an experimental equation and demonstrated the capability of the Potts model to provide information on microstructure for calculating the effective thermal conductivity of UO{sub 2} fuel.

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)

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

Three dimensions thermal-mechanical model of the billet in continuous casting petal-like mould

International Nuclear Information System (INIS)

Li Jing; Wu Li; Cao Zhiqiang; Tingju, L; Wang Tongmin

2012-01-01

Petal-like mould is a novel mould which has been applied to the steel industry in recent years. The behavior of the petal-like billet in continuous casting mould plays an important role in designing mould. It is hard to be in situ measured during continuous casting, however, can be worked out by the way of numerical simulation. But the research about the model of the billet in petal-like mould is very little. A 3D finite-element model has been built to simulate the thermal and stress fields of the molten steel in petal-like mould in this paper. The dynamic thermal boundary condition and the effect of ferrostatic pressure have been considered in the model. The temperature and stress in the billet have been predicted by this model.

Evaluation of uranium dioxide thermal conductivity using molecular dynamics simulations

International Nuclear Information System (INIS)

Kim, Woongkee; Kaviany, Massoud; Shim, J. H.

2014-01-01

It can be extended to larger space, time scale and even real reactor situation with fission product as multi-scale formalism. Uranium dioxide is a fluorite structure with Fm3m space group. Since it is insulator, dominant heat carrier is phonon, rather than electrons. So, using equilibrium molecular dynamics (MD) simulation, we present the appropriate calculation parameters in MD simulation by calculating thermal conductivity and application of it to the thermal conductivity of polycrystal. In this work, we investigate thermal conductivity of uranium dioxide and optimize the parameters related to its process. In this process, called Green Kubo formula, there are two parameters i.e correlation length and sampling interval, which effect on ensemble integration in order to obtain thermal conductivity. Through several comparisons, long correlation length and short sampling interval give better results. Using this strategy, thermal conductivity of poly crystal is obtained and comparison with that of pure crystal is made. Thermal conductivity of poly crystal show lower value that that of pure crystal. In further study, we broaden the study to transport coefficient of radiation damaged structures using molecular dynamics. Although molecular dynamics is tools for treating microscopic scale, most macroscopic issues related to nuclear materials such as voids in fuel materials and weakened mechanical properties by radiation are based on microscopic basis. Thus, research on microscopic scale would be expanded in this field and many hidden mechanism in atomic scales will be revealed via both atomic scale simulations and experiments

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.

Heavy truck modeling for fuel consumption. Simulations and measurements

Energy Technology Data Exchange (ETDEWEB)

Sandberg, T.

2001-12-01

Fuel consumption for heavy trucks depends on many factors like roads, weather, and driver behavior that are hard for a manufacturer to influence. However, one design possibility is the power train configuration. Here a new simulation program for heavy trucks is created to find the configuration of the power train that gives the lowest fuel consumption for each transport task. For efficient simulations the model uses production code for speed and gear control, and it uses exchangeable data sets to allow simulation of the whole production range of engine types, on recorded road profiles from all over the world. Combined with a graphical user interface this application is called STARS (Scania Truck And Road Simulation). The forces of rolling resistance and air resistance in the model are validated through an experiment where the propeller shaft torque of a heavy truck is measured. It is found that the coefficient of rolling resistance is strongly dependent on tire temperature, not only on vehicle speed as expected. This led to the development of a new model for rolling resistance. The model includes the dynamic behavior of the tires and relates rolling resistance to tire temperature and vehicle speed. In another experiment the fuel consumption of a test truck in highway driving is measured. The altitude of the road is recorded with a barometer and used in the corresponding simulations. Despite of the limited accuracy of this equipment the simulation program manage to predict a level of fuel consumption only 2% lower than the real measurements. It is concluded that STARS is a good tool for predicting fuel consumption for trucks in highway driving and for comparing different power train configurations.

Structure and thermal properties of as-fabricated U-7Mo/Mg and U-10Mo/Mg low-enriched uranium research reactor fuels

Energy Technology Data Exchange (ETDEWEB)

Kulakov, Mykola, E-mail: mykola.kulakov@cnl.ca [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Saoudi, Mouna [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Piro, Markus H.A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Donaberger, Ronald L. [Canadian Neutron Beam Centre, Chalk River, ON K0J 1J0 Canada (Canada)

2017-02-15

Aluminum-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements (with a core uranium loading of 4.5 gU/cm{sup 3}) have been fabricated at the Canadian Nuclear Laboratories for experimental tests and ultimately for use in research and test reactors. In this study, the microstructure and phase composition of unirradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were analyzed using optical and scanning electron microscopy, and neutron powder diffraction. Thermal properties were characterized using a combination of experimental measurements and thermodynamic calculations. The thermal diffusivity was measured using the laser flash method. The temperature-dependent specific heat capacities were calculated based on the linear rule of mixture using the weight fraction of different crystalline phases and their specific heat capacity values taken from the literature. The thermal conductivity was then calculated using the measured thermal diffusivity, the measured density and the calculated specific heat capacity. The resulting thermal conductivity is practically identical for both types of fuel. The in-reactor temperatures were predicted using conjugate heat transfer simulations. - Highlights: • Neutron diffraction analysis shows that most of the γ-U(Mo) phase was retained in as-fabricated U-7Mo/Mg and U-10Mo/Mg fuel cores. • The experimental thermal conductivity of both types of fuel is practically identical. • Based on conjugate heat transfer simulations, under normal operating conditions, the in-reactor fuel centreline temperature is about 510 K.

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)

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.)

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

Science.gov (United States)

Emrich, William J., Jr.

2014-01-01

Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit

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

Dynamic Simulations of Advanced Fuel Cycles

International Nuclear Information System (INIS)

Piet, Steven J.; Dixon, Brent W.; Jacobson, Jacob J.; Matthern, Gretchen E.; Shropshire, David E.

2011-01-01

Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the U.S. Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe 'lessons learned' from dynamic simulations but attempt to answer the 'so what' question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

Computer simulation of fuel element performance

Energy Technology Data Exchange (ETDEWEB)

Sukhanov, G I

1979-01-01

The review presents reports made at the Conference on the Bahaviour and Production of Fuel for Water Reactors on March 13-17, 1979. Discussed at the Conference are the most developed and tested calculation models specially evolved to predict the behaviour of fuel elements of water reactors. The following five main aspects of the problem are discussed: general conceptions and programs; mechanical mock-ups and their applications; gas release, gap conductivity and fuel thermal conductivity; analysis of nonstationary processes; models of specific phenomena. The review briefly describes the physical principles of the following models and programs: the RESTR, providing calculation of the radii of zones of columnar and equiaxial grains as well as the radius of the internal cavity of the fuel core; programs for calculation of fuel-can interaction, based on the finite elements method; a model predicting the behaviour of the CANDU-PHW fuel elements in transient conditions. General results are presented of investigations of heat transfer through a can-fuel gap and thermal conductivity of UO/sub 2/ with regard for cracking and gas release of the fuel. Many programs already suit the accepted standards and are intensively tested at present.

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

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

International Nuclear Information System (INIS)

McCulloch, R.W.; Post, D.W.; Lovell, R.T.; Snyder, S.D.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators

Simulation of leaking fuel rods

International Nuclear Information System (INIS)

Hozer, Z.

2006-01-01

The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123 I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123 I release from failed fuel rods during transients

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

International Nuclear Information System (INIS)

Yoo, Seung Hun

2011-02-01

The experimental and numerical studies of interim storages for nuclear spent fuels have been performed to investigate thermal-hydraulic characteristics of the dry storage systems and to propose new methodologies for the analysis and the design. Three separate researches have been performed in the present study: (a) Development of a scaling methodology and thermal-hydraulic experiment of a single spent fuel assembly simulating a dry storage cask: (b) Full-scope simulation of a dry storage cask by the use of Computational Fluid Dynamics (CFD) code: (c) Thermal-hydraulic design of a tunnel-type interim storage facility. In the first study, a scaling methodology has been developed to design a scaled-down canister. The scaling was performed in two steps. For the first step, the height of a spent fuel assembly was reduced from full height to half height. In order to consider the effect of height reduction on the natural convection, the scaling law of Ishii and Kataoka (1984) was employed. For the second step, the quantity of spent fuel assemblies was reduced from multiple assemblies to a single assembly. The scaling methodology was validated through the comparison with the experiment of the TN24P cask. The Peak Cladding Temperature (PCT), temperature gradients, and the axial and radial temperature distribution in the nondimensional forms were in good agreement with the experimental data. Based on the developed methodology, we have performed a single assembly experiment which was designed to simulate the full scale of the TN24P cask. The experimental data was compared with the CFD calculations. It turns out that their PCTs were less than the maximum allowable temperature for the fuel cladding and that the differences of their PCTs were agreed within 3 .deg. C, which was less than measurement uncertainty. In the second study, the full-scope simulations of the TN24P cask were performed by FLUENT. In order to investigate the sensitivity of the numerical and physical

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

ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation

International Nuclear Information System (INIS)

Parzer, I.; Kljenak, I.

2005-01-01

The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)

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

Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

International Nuclear Information System (INIS)

Simanullang, Irwan Liapto; Obara, Toru

2017-01-01

Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

Development of joining techniques for fabrication of fuel rod simulators

International Nuclear Information System (INIS)

Moorhead, A.J.; McCulloch, R.W.; Reed, R.W.; Woodhouse, J.J.

1980-10-01

Much of the safety-related thermal-hydraulic tests on nuclear reactors are conducted not in the reactor itself, but in mockup segments of a core that uses resistance-heated fuel rod simulators (FRS) in place of the radioactive fuel rods. Laser welding and furnace brazing techniques are described for joining subassemblies for FRS that have survived up to 1000 h steady-state operation at 700 to 1100 0 C cladding temperatures and over 5000 thermal transients, ranging from 10 to 100 0 C/s. A pulsed-laser welding procedure that includes use of small-diameter filler wire is used to join one end of a resistance heating element of Pt-8 W, Fe-22 Cr-5.5 Al-0.5 Co, or 80 Ni-20 Cr (wt %) to a tubular conductor of an appropriate intermediate material. The other end of the heating element is laser welded to an end plug, which in turn is welded to a central conductor rod

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.

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

Acoustic emission from fuel pellets in a simulated reactor environment

International Nuclear Information System (INIS)

Kupperman, D.S.; Kennedy, C.R.; Reimann, K.J.

1977-01-01

Thermal-shock damage of nuclear reactor fuel pellets in a simulated reactor environment has been correlated with acoustic-emission data obtained from sensors placed on extensions of the electrical feedthroughs. Ringdown counts, rms output data, and event-location data has been acquired for experiments carried out with single pellets as well as multiple pellet stacks. These tests have shown that acoustic-emission monitoring can provide information indicating the onset and the extent of cracking

User's manual for computer code SOLTES-1 (simulator of large thermal energy systems)

International Nuclear Information System (INIS)

Fewell, M.E.; Grandjean, N.R.; Dunn, J.C.; Edenburn, M.W.

1978-09-01

SOLTES simulates the steady-state response of thermal energy systems to time-varying data such as weather and loads. Thermal energy system models of both simple and complex systems can easily be modularly constructed from a library of routines. These routines mathematically model solar collectors, pumps, switches, thermal energy storage, thermal boilers, auxiliary boilers, heat exchangers, extraction turbines, extraction turbine/generators, condensers, regenerative heaters, air conditioners, heating and cooling of buildings, process vapor, etc.; SOLTES also allows user-supplied routines. The analyst need only specify fluid names to obtain readout of property data for heat-transfer fluids and constants that characterize power-cycle working fluids from a fluid property data bank. A load management capability allows SOLTES to simulate total energy systems that simultaneously follow heat and power loads and demands. Generalized energy accounting is available, and values for system performance parameters may be automatically determined by SOLTES. Because of its modularity and flexibility, SOLTES can be used to simulate a wide variety of thermal energy systems such as solar power/total energy, fossil fuel power plants/total energy, nuclear power plants/total energy, solar energy heating and cooling, geothermal energy, and solar hot water heaters

Development of variable width ribbon heating elements for liquid metal and gas-cooled fast breeder reactor fuel rod simulators

International Nuclear Information System (INIS)

McCulloch, R.W.; Lovell, R.T.; Post, D.W.; Snyder, S.D.

1980-01-01

Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod simulators used in test loops by the Breeder Reactor Program Thermal Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor Core Flow Test Loop. Thermal, mechanical, and electrical design considerations result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators

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)

International symposium on fuel rod simulators: development and application

Energy Technology Data Exchange (ETDEWEB)

McCulloch, R.W. (comp.)

1981-05-01

Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

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

Development of a numerical experimentation method for thermal hydraulics design and evaluation of high burn-up and innovative fuel pins

International Nuclear Information System (INIS)

Ninokata, Hisashi; Misawa, Takeharu; Baglietto, Emilio; Sorokin, A.P.; Maekawa, Isamu; Ohshima, Hiroyuki; Yamaguchi, Akira

2003-03-01

A method of large scale direct numerical simulation of turbulent flows in a high burn-up fuel pin bundle is proposed to evaluate wall shear stress and temperature distributions on the pin surfaces as well as detailed coolant velocity and temperature distributions inside subchannels under various thermal hydraulic conditions. This simulation is aimed at providing a tool to confirm margins to thermal hydraulics design limits of the nuclear fuels and at the same time to be used in design-by-analysis approaches. The method will facilitate thermal hydraulic design of high performance LMFR core fuels characterized by high burn-up, ultra long life, high reliable and safe performances, easiness of operation and maintenance, minimization of radio active wastes, without much relying on such empirical approach as hot spot factor and sub-factors, and above all the high cost mock up experiments. A pseudo direct numerical simulation of turbulence (DNS) code is developed, first on the Cartesian coordinates and then on the curvilinear boundary fit coordinates that enables us to reproduce thermal hydraulics phenomena in such a complicated flow channel as subchannels in a nuclear fuel pin assembly. The coordinate transformation is evaluated and demonstrated to yield correct physical quantities by carrying out computations and comparisons with experimental data with respect to the distributions of various physical quantities and turbulence statistics for fluid flow and heat transfers in various kinds of simple flow channel geometry. Then the boundary fitted pseudo DNS for flows inside an infinite pin array configuration is carried out and compared with available detailed experimental data. In parallel similar calculations are carried out using a commercial code STAR-CD to cross-check the DNS performances. As a results, the pseudo DNS showed reasonable comparisons with experiments as well as the STAR-CD results. Importance of the secondary flow influences is emphasized on the momentum

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.

Thermal Simulation of the Component Rework Profile Temperature

OpenAIRE

Nurminen, Janne

2015-01-01

The aim of this study was to clarify the possibilities and feasibility of the ther-mal simulation for the modeling of the rework process. The rework process modeling could enable an easy and fast access to the component and PWB level thermally critical effects like over and under heating of the component during the rework process. The modeling could also be used as a help of the real rework profile definition at an early phase of the electrical device development. The work includes a...

Direct numerical simulations of the ignition of lean primary reference fuel/air mixtures with temperature inhomogeneities

KAUST Repository

Luong, Minhbau

2013-10-01

The effects of fuel composition, thermal stratification, and turbulence on the ignition of lean homogeneous primary reference fuel (PRF)/air mixtures under the conditions of constant volume and elevated pressure are investigated by direct numerical simulations (DNSs) with a new 116-species reduced kinetic mechanism. Two-dimensional DNSs were performed in a fixed volume with a two-dimensional isotropic velocity spectrum and temperature fluctuations superimposed on the initial scalar fields with different fuel compositions to elucidate the influence of variations in the initial temperature fluctuation and turbulence intensity on the ignition of three different lean PRF/air mixtures. In general, it was found that the mean heat release rate increases slowly and the overall combustion occurs fast with increasing thermal stratification regardless of the fuel composition under elevated pressure and temperature conditions. In addition, the effect of the fuel composition on the ignition characteristics of PRF/air mixtures was found to vanish with increasing thermal stratification. Chemical explosive mode (CEM), displacement speed, and Damköhler number analyses revealed that the high degree of thermal stratification induces deflagration rather than spontaneous ignition at the reaction fronts, rendering the mean heat release rate more distributed over time subsequent to thermal runaway occurring at the highest temperature regions in the domain. These analyses also revealed that the vanishing of the fuel effect under the high degree of thermal stratification is caused by the nearly identical propagation characteristics of deflagrations of different PRF/air mixtures. It was also found that high intensity and short-timescale turbulence can effectively homogenize mixtures such that the overall ignition is apt to occur by spontaneous ignition. These results suggest that large thermal stratification leads to smooth operation of homogeneous charge compression-ignition (HCCI

Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

OpenAIRE

Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

2016-01-01

Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

FUDA MOD-2: a computer program for simulation the performance of fuel element validation exercise

International Nuclear Information System (INIS)

Chouhan, S.K.; Tripathi, R.M.; Prasad, P.N.; Chauhan, Ashok

2014-01-01

The PHWR fuel element performance is evaluated using the fuel analysis computer code FUDA MOD2. It is specifically written for performance simulation of UO 2 fuel pellet, located in zirconium alloy sheath operating under coolant pressure. For specific element power histories, the code investigates the variables and their interactions that govern fuel element performance. The input data requires pellet dimensions, element dimensions, sheath properties, heat transfer data, thermal hydraulic parameters of coolant, the inner filler gas composition, flux gradient and linear heat ratings (LHR) at different burn up. The output data generated by the code are radial temperature profile of fuel and sheath, fuel sheath-gap heat transfer coefficient, fission gas generated and released, fission gas pressure, sheath stress and strain for different burn-up zones. The code has been verified against literature data and post irradiation examinations carried out. It is also bench marked against various international fuel element simulation programmes available with water cooled reactors operating countries. The present paper describes the FUDA MOD2 code verification studies carried out using the literature data and post irradiation examination data. (author)

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

Minerve: thermal-hydraulic phenomena simulation and virtual reality

International Nuclear Information System (INIS)

Laffont, A.; Pentori, B.

2003-01-01

MINERVE is a 3D interactive application representing the thermal-hydraulic phenomena happening in a nuclear plant. Therefore, the 3D geometric model of the French 900 MW PWR installations has been built. The users can interact in real time with this model to see at each step of the simulation what happens in the pipes. The thermal-hydraulic simulation is made by CATHARE-2, which calculates at every time step data on about one thousand meshes (the whole primary circuit, a part of the second circuit, and the Residual Heat Removal System). The simulation covers incidental and accidental cases on these systems. There are two main innovations in MINERVE: In the domain of nuclear plant's visualization, it is to introduce interactive 3D software mechanisms to visualize results of a physical simulation. In the domain of real-time 3D, it is to visualize fluids in a pipe, while they can have several configurations, like bubbles or single liquid phase. These mechanisms enable better comprehension and better visual representation of the possible phenomena. This paper describes the functionalities of MINERVE, and the difficulties to represent fluids with several characteristics like speed, configuration,..., in 3D. On the end, we talk about the future of MINERVE, and more widely of the possible futures of such an application in scientific visualization. (authors)

Minerve: thermal-hydraulic phenomena simulation and virtual reality

Energy Technology Data Exchange (ETDEWEB)

Laffont, A.; Pentori, B. [EDF R and D, EDF SEPTEN Electricity of France - Research and Development, Department SINETICS, 92 - Clamart (France)

2003-07-01

MINERVE is a 3D interactive application representing the thermal-hydraulic phenomena happening in a nuclear plant. Therefore, the 3D geometric model of the French 900 MW PWR installations has been built. The users can interact in real time with this model to see at each step of the simulation what happens in the pipes. The thermal-hydraulic simulation is made by CATHARE-2, which calculates at every time step data on about one thousand meshes (the whole primary circuit, a part of the second circuit, and the Residual Heat Removal System). The simulation covers incidental and accidental cases on these systems. There are two main innovations in MINERVE: In the domain of nuclear plant's visualization, it is to introduce interactive 3D software mechanisms to visualize results of a physical simulation. In the domain of real-time 3D, it is to visualize fluids in a pipe, while they can have several configurations, like bubbles or single liquid phase. These mechanisms enable better comprehension and better visual representation of the possible phenomena. This paper describes the functionalities of MINERVE, and the difficulties to represent fluids with several characteristics like speed, configuration,..., in 3D. On the end, we talk about the future of MINERVE, and more widely of the possible futures of such an application in scientific visualization. (authors)

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)

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.

Thermal expansion

International Nuclear Information System (INIS)

Yun, Y.

2015-01-01

Thermal expansion of fuel pellet is an important property which limits the lifetime of the fuels in reactors, because it affects both the pellet and cladding mechanical interaction and the gap conductivity. By fitting a number of available measured data, recommended equations have been presented and successfully used to estimate thermal expansion coefficient of the nuclear fuel pellet. However, due to large scatter of the measured data, non-consensus data have been omitted in formulating the equations. Also, the equation is strongly governed by the lack of appropriate experimental data. For those reasons, it is important to develop theoretical methodologies to better describe thermal expansion behaviour of nuclear fuel. In particular, first-principles and molecular dynamics simulations have been certainly contributed to predict reliable thermal expansion without fitting the measured data. Furthermore, the two theoretical techniques have improved on understanding the change of fuel dimension by describing the atomic-scale processes associated with lattice expansion in the fuels. (author)

Weight simulation fuel assembly

International Nuclear Information System (INIS)

Sumikawa, Kiyokazu; Tokomatsu, Mutsuo.

1993-01-01

A tungsten pellet is not applied with hollow fabrication but a tungsten rod is inserted and filled into a zircaloy fuel cladding tube, as well as different kind of material having a density lower than that of tungsten, for example, stainless steel rods, are disposed successively intermittently and alternately for simulating the weight of one fuel rod. The filling method and the length of the individual pellets are optional depending on the method of usage, providing that the outer diameter of the simulation pellet is made identical with that of the actual fuel pellet. With such a constitution, there is no need to dispose a hollow portion as in the case of using only tungsten pellets, and the costs for both the materials and the fabrication can be saved. (T.M.)

Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

International Nuclear Information System (INIS)

EL Khatib, H.H.A.

2013-01-01

The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

Fission gas induced deformation model for FRAP-T6 and NSRR irradiated fuel test simulations

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hosoyamada, Ryuji; Mori, Yukihide

1996-11-01

Pulse irradiation tests of irradiated fuels under simulated reactivity initiated accidents (RIAs) have been carried out at the Nuclear Safety Research Reactor (NSRR). Larger cladding diameter increase was observed in the irradiated fuel tests than in the previous fresh fuel tests. A fission gas induced cladding deformation model was developed and installed in a fuel behavior analysis code, FRAP-T6. The irradiated fuel tests were analyzed with the model in combination with modified material properties and fuel cracking models. In Test JM-4, where the cladding temperature rose to higher temperatures and grain boundary separation by the pulse irradiation was significant, the fission gas model described the cladding deformation reasonably well. The fuel had relatively flat radial power distribution and the grain boundary gas from the whole radius was calculated to contribute to the deformation. On the other hand, the power density in the irradiated LWR fuel rods in the pulse irradiation tests was remarkably higher at the fuel periphery than the center. A fuel thermal expansion model, GAPCON, which took account of the effect of fuel cracking by the temperature profile, was found to reproduce well the LWR fuel behavior with the fission gas deformation model. This report present details of the models and their NSRR test simulations. (author)

Thermodynamic analysis of fuels in gas phase: ethanol, gasoline and ethanol - gasoline predicted by DFT method.

Science.gov (United States)

Neto, A F G; Lopes, F S; Carvalho, E V; Huda, M N; Neto, A M J C; Machado, N T

2015-10-01

This paper presents a theoretical study using density functional theory to calculate thermodynamics properties of major molecules compounds at gas phase of fuels like gasoline, ethanol, and gasoline-ethanol mixture in thermal equilibrium on temperature range up to 1500 K. We simulated a composition of gasoline mixture with ethanol for a thorough study of thermal energy, enthalpy, Gibbs free energy, entropy, heat capacity at constant pressure with respect to temperature in order to study the influence caused by ethanol as an additive to gasoline. We used semi-empirical computational methods as well in order to know the efficiency of other methods to simulate fuels through this methodology. In addition, the ethanol influence through the changes in percentage fractions of chemical energy released in combustion reaction and the variations on thermal properties for autoignition temperatures of fuels was analyzed. We verified how ethanol reduces the chemical energy released by gasoline combustion and how at low temperatures the gas phase fuels in thermal equilibrium have similar thermodynamic behavior. Theoretical results were compared with experimental data, when available, and showed agreement. Graphical Abstract Thermodynamic analysis of fuels in gas phase.

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

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

Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies

International Nuclear Information System (INIS)

Rose, S.D.; Carbajo, J.J.; Levin, A.E.; Lloyd, D.B.; Montgomery, B.H.; Wantland, J.L.

1985-01-01

Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal 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.)

Nuclear fuel cycle simulation system (VISTA)

International Nuclear Information System (INIS)

2007-02-01

The Nuclear Fuel Cycle Simulation System (VISTA) is a simulation system which estimates long term nuclear fuel cycle material and service requirements as well as the material arising from the operation of nuclear fuel cycle facilities and nuclear power reactors. The VISTA model needs isotopic composition of spent nuclear fuel in order to make estimations of the material arisings from the nuclear reactor operation. For this purpose, in accordance with the requirements of the VISTA code, a new module called Calculating Actinide Inventory (CAIN) was developed. CAIN is a simple fuel depletion model which requires a small number of input parameters and gives results in a very short time. VISTA has been used internally by the IAEA for the estimation of: spent fuel discharge from the reactors worldwide, Pu accumulation in the discharged spent fuel, minor actinides (MA) accumulation in the spent fuel, and in the high level waste (HLW) since its development. The IAEA decided to disseminate the VISTA tool to Member States using internet capabilities in 2003. The improvement and expansion of the simulation code and the development of the internet version was started in 2004. A website was developed to introduce the simulation system to the visitors providing a simple nuclear material flow calculation tool. This website has been made available to Member States in 2005. The development work for the full internet version is expected to be fully available to the interested parties from IAEA Member States in 2007 on its website. This publication is the accompanying text which gives details of the modelling and an example scenario

The Effect of Fuel Mass Fraction on the Combustion and Fluid Flow in a Sulfur Recovery Unit Thermal Reactor

Directory of Open Access Journals (Sweden)

Chun-Lang Yeh

2016-11-01

Full Text Available Sulfur recovery unit (SRU thermal reactors are negatively affected by high temperature operation. In this paper, the effect of the fuel mass fraction on the combustion and fluid flow in a SRU thermal reactor is investigated numerically. Practical operating conditions for a petrochemical corporation in Taiwan are used as the design conditions for the discussion. The simulation results show that the present design condition is a fuel-rich (or air-lean condition and gives acceptable sulfur recovery, hydrogen sulfide (H2S destruction, sulfur dioxide (SO2 emissions and thermal reactor temperature for an oxygen-normal operation. However, for an oxygen-rich operation, the local maximum temperature exceeds the suggested maximum service temperature, although the average temperature is acceptable. The high temperature region must be inspected very carefully during the annual maintenance period if there are oxygen-rich operations. If the fuel mass fraction to the zone ahead of the choke ring (zone 1 is 0.0625 or 0.125, the average temperature in the zone behind the choke ring (zone 2 is higher than the zone 1 average temperature, which can damage the downstream heat exchanger tubes. If the zone 1 fuel mass fraction is reduced to ensure a lower zone 1 temperature, the temperature in zone 2 and the heat exchanger section must be monitored closely and the zone 2 wall and heat exchanger tubes must be inspected very carefully during the annual maintenance period. To determine a suitable fuel mass fraction for operation, a detailed numerical simulation should be performed first to find the stoichiometric fuel mass fraction which produces the most complete combustion and the highest temperature. This stoichiometric fuel mass fraction should be avoided because the high temperature could damage the zone 1 corner or the choke ring. A higher fuel mass fraction (i.e., fuel-rich or air-lean condition is more suitable because it can avoid deteriorations of both zone 1

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.

A parallel multi-domain solution methodology applied to nonlinear thermal transport problems in nuclear fuel pins

Energy Technology Data Exchange (ETDEWEB)

Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Berrill, Mark A.; Allu, Srikanth; Hamilton, Steven P.; Sampath, Rahul S.; Clarno, Kevin T. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Dilts, Gary A. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)

2015-04-01

This paper describes an efficient and nonlinearly consistent parallel solution methodology for solving coupled nonlinear thermal transport problems that occur in nuclear reactor applications over hundreds of individual 3D physical subdomains. Efficiency is obtained by leveraging knowledge of the physical domains, the physics on individual domains, and the couplings between them for preconditioning within a Jacobian Free Newton Krylov method. Details of the computational infrastructure that enabled this work, namely the open source Advanced Multi-Physics (AMP) package developed by the authors is described. Details of verification and validation experiments, and parallel performance analysis in weak and strong scaling studies demonstrating the achieved efficiency of the algorithm are presented. Furthermore, numerical experiments demonstrate that the preconditioner developed is independent of the number of fuel subdomains in a fuel rod, which is particularly important when simulating different types of fuel rods. Finally, we demonstrate the power of the coupling methodology by considering problems with couplings between surface and volume physics and coupling of nonlinear thermal transport in fuel rods to an external radiation transport code.

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

Thermal characteristics during hydrogen fueling process of type IV cylinder

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Chan [Department of Fire and Disaster Prevention, Kyungil University, 33, Buhori, Hayang, Kyungsan 712-701 (Korea); Lee, Seung Hoon; Yoon, Kee Bong [Department of Mechanical Engineering, Chung Ang University, 221, Huksuk, Dongjak, Seoul 156-756 (Korea)

2010-07-15

Temperature increase during hydrogen fueling process is a significant safety concern of a high pressure hydrogen vessel. Hence, thermal characteristics of a Type IV cylinder during hydrogen filling process need to be understood. In this study, a series of experiments were conducted to quantify the temperature change of the cylinder during hydrogen filling to 35 MPa. Computational fluid dynamics (CFD) analysis was also conducted to simulate the conditions of the experiments. The results predicted by the CFD analysis show reasonable agreement with the experiments and the discrepancy between the CFD results and experimental results decrease with higher initial gas pressures. The upper and the lower parts of the vessel showed a temperature difference in the vertical direction. The upper gas temperature was higher than that of the lower part due to the buoyancy effect in the vessel. The maximum gas temperature was higher than the maximum temperature allowed in the ISO safety code (85 C) for the case in which the vessel was pressurized from 0 MPa to 35 MPa. This work contributes to the understanding of the thermal flow characteristics of the hydrogen filling process and notes that additional efforts should be made to guarantee the safety of a type IV cylinder during the hydrogen fueling process. (author)

Thermal-Hydraulics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

Energy Technology Data Exchange (ETDEWEB)

Bajorek, Stephen; Diamond, David J.

2018-11-11

This paper discusses liquid-fuel molten salt reactors, how they will operate under normal, transient, and accident conditions, and the results of an expert elicitation to determine the corresponding thermalhydraulic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC) to develop or identify modeling functionalities and tools required to carry out confirmatory analyses that examine the validity and accuracy of an applicant’s calculations and help determine the margin of safety in plant design. NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures-of-merit (FoMs) (e.g., salt temperature, velocity, and composition). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs.

Innovative nuclear fuels: results and strategy

International Nuclear Information System (INIS)

Stan, Marius

2009-01-01

To facilitate the discovery and design of innovative nuclear fuels, multi-scale models and simulations are used to predict irradiation effects on the thermal conductivity, oxygen diffusivity, and thermal expansion of oxide fuels. The multi-scale approach is illustrated using results on ceramic fuels with a focus on predictions of point defect concentrations, stoichiometry, and phase stability. The high performance computer simulations include coupled heat transport, diffusion, and thermal expansion, gas bubble formation and temperature evolution in a fuel element consisting of UO2 fuel and metallic cladding. The second part of the talk is dedicated to a discussion of an international strategy for developing advanced, innovative nuclear fuels. Four initiative are proposed to accelerate the discovery and design of new materials: (a) Develop an international pool of experts, (b) Create Institutes for Materials Discovery and Design, (c) Create an International Knowledge base for experimental data, models (mathematical expressions), and simulations (codes) and (d) Organize international workshops and conference sessions. The paper ends with a discussion of existing and emerging international collaborations.

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)

Modeling of the thermo-mechanical behaviour of the PWR fuel

International Nuclear Information System (INIS)

Mailhe, P.

2014-01-01

This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

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-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

Directory of Open Access Journals (Sweden)

Itamar Iliuk

2016-01-01

Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

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

Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

Energy Technology Data Exchange (ETDEWEB)

Hartmann, Christoph Oliver

2016-06-13

Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

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

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

Simulation of the neutron-physical properties of the classical UO2 fuel and of MOX fuel during the burn-up by Transuranus

International Nuclear Information System (INIS)

Breza, J. jr.; Necas, V.; Daoeilek, P.

2005-01-01

The classical nuclear fuel UO 2 is well known for VVER reactors. Nevertheless, in the near future it will be possible to replace this fuel by novel, advanced kinds of fuel, for instance MOX, inert matrices fuel, etc., that will allow to increase the level of burn-up and minimize the amount of hazardous waste. The code Transuranus [2], designed at ITU Karlsruhe, is intended for thermal and mechanical analyses of fuel elements in nuclear reactors. We have utilized the code Transuranus to simulate the neutron-physical properties of the classical UO 2 fuel and of MOX fuel during the burn-up to a level of 40 MWd/kgHM. We compare obtained results of uranium and plutonium nuclides concentrations, their changes during burn-up, with results obtained by code HELIOS [3], which is well-validated code for this kind of applications. We performed calculations of fission gasses concentrations, namely xenon and krypton. (author)

Development of the fabrication technology of the simulated DUPIC fuel

International Nuclear Information System (INIS)

Kang, Kweon Ho; Yang, M. S.; Bae, K. K. and others

2000-06-01

It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties of the DUPIC fuel is different from the commercial UO 2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, processes on powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using simulated spent fuel are discribed. To fabricate simulated DUPIC fuel, the powder from 3 times OREOX and 5 times attrition milling simulated spent fuel is compacted with 1.3 ton/cm 2 . Pellets are sintered in 100% H 2 atmosphere over 10 h at 1800 deg C. Sintered densities of pellets are 10.2-10.5 g/cm 3

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

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)

Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

2003-04-01

A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

Thermal transport in UO₂ with defects and fission products by 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); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lashley, Jason Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Byler, Darrin David [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)

2015-10-14

The importance of the thermal transport in nuclear fuel has motivated a wide range of experimental and modelling studies. In this report, the reduction of thermal transport in UO₂ due to defects and fission products has been investigated using non-equilibrium MD simulations, with two sets of empirical potentials for studying the degregation of UO₂ thermal conductivity including a Buckingham type interatomic potential and a recently developed EAM type interatomic potential. Additional parameters for U⁵⁺ and Zr⁴⁺ in UO₂ have been developed for the EAM potential. The thermal conductivity results from MD simulations are then corrected for the spin-phonon scattering through Callaway model formulations. To validate the modelling results, comparison was made with experimental measurements on single crystal hyper-stoichiometric UO_2+x samples.

The fuel cell model of abiogenesis: a new approach to origin-of-life simulations.

Science.gov (United States)

Barge, Laura M; Kee, Terence P; Doloboff, Ivria J; Hampton, Joshua M P; Ismail, Mohammed; Pourkashanian, Mohamed; Zeytounian, John; Baum, Marc M; Moss, John A; Lin, Chung-Kuang; Kidd, Richard D; Kanik, Isik

2014-03-01

In this paper, we discuss how prebiotic geo-electrochemical systems can be modeled as a fuel cell and how laboratory simulations of the origin of life in general can benefit from this systems-led approach. As a specific example, the components of what we have termed the "prebiotic fuel cell" (PFC) that operates at a putative Hadean hydrothermal vent are detailed, and we used electrochemical analysis techniques and proton exchange membrane (PEM) fuel cell components to test the properties of this PFC and other geo-electrochemical systems, the results of which are reported here. The modular nature of fuel cells makes them ideal for creating geo-electrochemical reactors with which to simulate hydrothermal systems on wet rocky planets and characterize the energetic properties of the seafloor/hydrothermal interface. That electrochemical techniques should be applied to simulating the origin of life follows from the recognition of the fuel cell-like properties of prebiotic chemical systems and the earliest metabolisms. Conducting this type of laboratory simulation of the emergence of bioenergetics will not only be informative in the context of the origin of life on Earth but may help in understanding whether life might emerge in similar environments on other worlds.

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.)

Simulation of Missing Pellet Surface thermal behavior with 3D dynamic gap element

International Nuclear Information System (INIS)

Kim, Hyo Chan; Yang, Yong Sik; Koo, Yang Hyun; Kang, Chang Hak; Lee Sung Uk; Yang, Dong Yol

2014-01-01

Most of the fuel performance codes that are able to simulate a multidimensional analysis are used to calculate the radial temperature distribution and perform a multidimensional mechanical analysis based on a one-dimensional (1D) temperature result. The FRAPCON-FRAPTRAN code system incorporates a 1D thermal module and two-dimensional (2D) mechanical module when FEM option is activated. In this method, the multidimensional gap conductance model is not required because one-dimensional thermal analysis is carried out. On the other hand, a gap conductance model for a multi-dimension should be developed in the code to perform a multidimensional thermal analysis. ALCYONE developed by CEA introduces an equivalent heat convection coefficient that represents the multidimensional gap conductance. However, the code does not employ dynamic gap conductance which is a function of gap thickness and gap characteristics in direct. The BISON code, which has been developed by INL (Idaho National Laboratory), employed a thermo-mechanical contact method that is specifically designed for tightly-coupled implicit solutions that employ Jacobian-free solution methods. Owing to tightly-coupled implicit solutions, the BISON code solves gap conductance and gap thickness simultaneously with given boundary conditions. In this paper, 3D dynamic gap element has been proposed to resolve convergence issue and nonlinear characteristic of multidimensional gap conductance. To evaluate 3D dynamic gap element module, 3D thermomechanical module using FORTRAN77 has been implemented incorporating 3D dynamic gap element. To demonstrate effect of 3D dynamic gap element, thermal behavior of missing pellet surface (MPS) has been simulated by the developed module. LWR fuel performance codes should incorporate thermo-mechanical loop to solve gap conductance problem, iteratively. However, gap conductance in multidimensional model is difficult issue owing to its nonlinearity and convergence characteristics. In

Thermal analysis of HNPF spent fuel shipping container in torch environments

International Nuclear Information System (INIS)

Eggers, P.E.; Crawford, H.L.; Trujillo, A.A.; Pope, R.B.; Yoshimura, H.R.

1980-01-01

A thermal analysis of a modified HNPF cask has been performed to predict cask temperatures in simulated torch fire environments. A portion of this analysis also provided a basis for the modifications leading to the design of external cooling fins, a pressure relief valve system, a spent fuel torch fire analyses will provide comparison with experimentally measured temperatures of the HNPF cask, in torch fire environments, which will be obtained in the future. Scope of this study was analytical in nature and involved the use of a computer-based heat transfer analysis methods. Product of this study is the prediction of cask temperatures and internal pressures under normal and torch-fire accident conditions. This study also involved an analysis of the torch fire literature in order to develop an empirically based thermal model for the torch environment. 8 figures, 3 tables

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

Development of a syngas-fired catalytic combustion system for hybrid solar-thermal applications

International Nuclear Information System (INIS)

Gupta, Mayank; Pramanik, Santanu; Ravikrishna, R.V.

2016-01-01

Highlights: • Syngas-fired combustor concept as hybrid heat source for solar thermal application. • Experimental characterization of catalytic combustor under fuel-rich conditions. • Stable operation, quick startup, and high turn-down ratio demonstrated. • Reacting flow CFD simulations of single channel of catalytic monolith. - Abstract: This paper describes the development and operation of a catalytic combustion system for use with syngas as an important component of a hybrid heating source for solar-thermal power generation. The reactor consists of a cylindrical ceramic monolith with porous alumina washcoat in which platinum is distributed as the catalyst. Two fuel-rich equivalence ratios were studied over a range of flow rates. The fuel-rich conditions permit low temperature combustion without the problem of hotspots likely to occur under fuel-lean conditions with hydrogen-containing fuels. Experimental data of temperature and species concentration at the exit of the reactor have been reported for a maximum fuel thermal input of 34 kW. The system exhibited quick start-up with a light-off time of around 60 s and a steady-state time of around 200 s as determined from the transient temperature profiles. The experimental results have also been complemented with detailed two-dimensional numerical simulations for improved understanding of the combustion characteristics in the reactor. The simulations suggest that the combustion system can be operated at a turn-down ratios far in excess of 1.67, which is the maximum value that has been investigated in the present setup. Stable operation, quick startup, and high turn-down ratio are some of the key features that enable the proposed combustion system to accommodate the transients in solar-thermal applications.

Enhancing the ABAQUS Thermomechanics Code to Simulate Multidimensional Steady and Transient Fuel Rod Behavior

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L.; Knoll, D.A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-3855 (United States)

2009-06-15

isotropic and anisotropic elasticity, thermal expansion, plasticity, and thermal creep. Due to its commercial nature and widespread use, ABAQUS includes powerful pre- and post-processing software, facilitating efficient problem setup and rapid interpretation of results. The code undergoes extensive and well-documented verification and validation procedures, reducing the risk of coding errors leading to invalid results. Essential for the intended application, ABAQUS includes a user interface permitting the inclusion of user-developed models to simulate fuel-specific phenomena. This powerful interface makes it possible to define any constitutive model of arbitrary complexity. Focusing initially on UO{sub 2} fuel, user subroutines have been developed and tested to describe temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, and fission gas release. Assuming Zircaloy as the clad material, models have been included for both thermal and irradiation creep. A subroutine has been developed to model the complex heat exchange across the fuel-clad gap. This model accounts for the temperature, pressure, and species-dependent gap gas thermal conductivity, the changing gap width, the roughness of both the fuel and clad, jump distances, contact pressure, and the cladding hardness. Gap heat transfer is tightly coupled to the fission gas model and the basic mechanics. Field variables are defined at each integration point to track fuel specific phenomena such as burnup, fission gas produced, and fission gas released. Fuels modeling capability is demonstrated using a 2D axisymmetric analysis of a rod section including two individual pellets, the associated cladding, the gas-filled gap, and an upper plenum region. The plenum and gap gas are modeled as a 'hydro' cavity in ABAQUS with the cavity volume dictated by the evolving solid mechanics, the gas mass and species mixture, and the gas state computed assuming ideal gas behavior. A simple

Application of process simulation for evaluation of ecologically benefical developments in thermal power plant technology

International Nuclear Information System (INIS)

Schuster, G.

2000-04-01

Responsibility for the environment and a sustainable utilization of resources gain also in the production of electric power more and more importance. For this reason existing power generation processes have to be improved and alternatives to existing processes have to be developed. As a first step in this procedure process simulation is a powerful tool to evaluate the potentials of new developments. In this work it is shown, how new thermal power processes are modeled and simulated based on well-known thermodynamic and chemical correlations. Processes for thermal power plants using lignite with high water content and biomass as fuel are studied. In each case simulations are carried out for complete plants including all important unit operations. Based on a conventional thermal power plant for lignite different variants for efficiency improvement by fuel drying are examined. Additionally the potential of a process with gasification and gas turbine is discussed. Compared to a lignite power plant the preconditions for a biomass power plant are different. A promising option for the future seems to be small, decentralized combined heat and power plants. Therefore a process with simple and compact design including gasifier and gas turbine is regarded and sensitivity analyses are carried out. As well as for the lignite processes possible improvements by fuel drying are studied. The basis lignite power plant (drying in an impact rotor mill with hot flue gas) has an overall electric efficiency of 36 %. Alternative fuel drying processes (reducing water content from 54 w % to 10 w %) can increase efficiency to nearly 43 %. Using integrated air-blown gasification combined with gas turbine and steam turbine and additional fuel drying raises the efficiency up to 49 % in the case of cold gas cleanup and up to 50 percent in the case of hot gas cleanup. Efficiencies of the regarded biomass power plants are in the range of about 20 % (with a biomass water content of 25 w %). By

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)

GAPCON-THERMAL-3

International Nuclear Information System (INIS)

Mohr, C.L.; Lanning, D.D.; Panisko, F.E.

1979-01-01

The fuel performance code GAPCON-THERMAL-3 has been expanded to include recent transient material deformation constitutive relations and the FLECHT heat transfer correlation. The modifications make it possible to compute the thermal and mechanical response of nuclear fuel to postulated Loss of Coolant Accidents (LOCA). The numerical formulation has the capability of predicting both steady state and transient behavior of a fuel rod using a single analytical procedure. GAPCON-THERMAL-3 (G-T-3) uses a specialized finite element procedure for mechanics predictions and the method of weighted residuals and finite difference techniques to compute temperature and thermal behavior. Fuel behavior, gas release models, gas conductance models, and stored energy calculations are applicable to both steady state and transient conditions. The code has been used to perform scoping analysis for in-reactor LOCA simulation testing. (orig.)

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.

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

International Nuclear Information System (INIS)

Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-01

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

Energy Technology Data Exchange (ETDEWEB)

Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-09

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Directory of Open Access Journals (Sweden)

Skrzypek Maciej

2015-09-01

Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

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)

Fuel lattice design using heuristics and new strategies

Energy Technology Data Exchange (ETDEWEB)

Ortiz S, J. J.; Castillo M, J. A.; Torres V, M.; Perusquia del Cueto, R. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico); Pelta, D. A. [ETS Ingenieria Informatica y Telecomunicaciones, Universidad de Granada, Daniel Saucedo Aranda s/n, 18071 Granada (Spain); Campos S, Y., E-mail: juanjose.ortiz@inin.gob.m [IPN, Escuela Superior de Fisica y Matematicas, Unidad Profesional Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

2010-10-15

This work show some results of the fuel lattice design in BWRs when some allocation pin rod rules are not taking into account. Heuristics techniques like Path Re linking and Greedy to design fuel lattices were used. The scope of this work is to search about how do classical rules in design fuel lattices affect the heuristics techniques results and the fuel lattice quality. The fuel lattices quality is measured by Power Peaking Factor and Infinite Multiplication Factor at the beginning of the fuel lattice life. CASMO-4 code to calculate these parameters was used. The analyzed rules are the following: pin rods with lowest uranium enrichment are only allocated in the fuel lattice corner, and pin rods with gadolinium cannot allocated in the fuel lattice edge. Fuel lattices with and without gadolinium in the main diagonal were studied. Some fuel lattices were simulated in an equilibrium cycle fuel reload, using Simulate-3 to verify their performance. So, the effective multiplication factor and thermal limits can be verified. The obtained results show a good performance in some fuel lattices designed, even thought, the knowing rules were not implemented. A fuel lattice performance and fuel lattice design characteristics analysis was made. To the realized tests, a dell workstation was used, under Li nux platform. (Author)

Fuel lattice design using heuristics and new strategies

International Nuclear Information System (INIS)

Ortiz S, J. J.; Castillo M, J. A.; Torres V, M.; Perusquia del Cueto, R.; Pelta, D. A.; Campos S, Y.

2010-10-01

This work show some results of the fuel lattice design in BWRs when some allocation pin rod rules are not taking into account. Heuristics techniques like Path Re linking and Greedy to design fuel lattices were used. The scope of this work is to search about how do classical rules in design fuel lattices affect the heuristics techniques results and the fuel lattice quality. The fuel lattices quality is measured by Power Peaking Factor and Infinite Multiplication Factor at the beginning of the fuel lattice life. CASMO-4 code to calculate these parameters was used. The analyzed rules are the following: pin rods with lowest uranium enrichment are only allocated in the fuel lattice corner, and pin rods with gadolinium cannot allocated in the fuel lattice edge. Fuel lattices with and without gadolinium in the main diagonal were studied. Some fuel lattices were simulated in an equilibrium cycle fuel reload, using Simulate-3 to verify their performance. So, the effective multiplication factor and thermal limits can be verified. The obtained results show a good performance in some fuel lattices designed, even thought, the knowing rules were not implemented. A fuel lattice performance and fuel lattice design characteristics analysis was made. To the realized tests, a dell workstation was used, under Li nux platform. (Author)

A proposal for a unified fuel thermal conductivity model available for UO{sub 2}, (U-Pu)O{sub 2} and UO{sub 2}-GD{sub 2}O{sub 3} PWR fuel

Energy Technology Data Exchange (ETDEWEB)

Baron, D [Electrice de France, Moret-sur-Loing (France); Couty, J C [Electricite de France (EDF), 69 - Villeurbanne (France)

1997-08-01

In order to cope with the current fuel management targets which are focussed on higher discharge burnups, initial {sup 235}U fuel enrichments have been increased from 3.25% to 4%. To avoid an increase in boron concentration in the primary circuit, Gadolinium is used as a burnable poison, spread in the uranium oxide matrix of selected rods, in order to absorb the initial reactivity excess. Obviously, fuel thermal conductivity is affected when introducing any stranger element. Previously, the EDF thermomechanical code provided two different models to simulate the fuel thermal conductivity: one available for UO{sub 2} and (U-Pu)O{sub 2} fuels, the other for Gadolinia fuels, depending on the calculations to be done. No effect of the initial fuel stoichiometry was taken into account in the second model. That situation suggested the development of a unified model available for any fuels presently loaded in the EDF PWR reactors. This paper deals with the choice of the formulation, the data base used and the methodology applied for parameter fitting. Results in terms of measured versus predicted evaluation are then discussed. (author). 11 refs, 5 figs.

Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

Energy Technology Data Exchange (ETDEWEB)

Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

2017-04-15

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

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.)

Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment

Energy Technology Data Exchange (ETDEWEB)

Quastel, A.; Thiriet, C. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Lewis, B., E-mail: brent.lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada); Corcoran, E., E-mail: emily.corcoran@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

An experimental program funded by the CANDU Owners Group (COG) is studying an out-reactor instrumented defected fuel experiment in Stern Laboratories (Hamilton, Ontario) with guidance from Atomic Energy of Canada Limited (AECL). The objective of this test is to provide experimental data for validation of a mechanistic fuel oxidation model. In this experiment a defected fuel element with UO{sub 2} pellets will be internally heated with an electrical heater element, causing the fuel to crack. By defecting the sheath in-situ the fuel will be exposed to light water coolant near normal reactor operating conditions (pressure 10 MPa and temperature 265-310{sup o}C) causing fuel oxidation, especially near the hotter regions of the fuel in the cracks. The fuel thermal conductivity will change, resulting in a change in the temperature distribution of the fuel element. This paper provides 2D r-θ plane strain solid mechanics models to simulate fuel thermal expansion, where conditions for fuel crack propagation are investigated with the thermal J integral to predict fuel crack stress intensity factors. Finally since fuel crack geometry can affect fuel oxidation this paper shows that the solid mechanics model with pre-set radial cracks can be coupled to a 2D r-θ fuel oxidation model. (author)

Computer Simulations of Composite Electrodes in Solid-Oxide Fuel-Cells

Energy Technology Data Exchange (ETDEWEB)

Sunde, Svein

1999-07-01

Fuel cells are devices for converting the combined chemical (free) energy of fuels and oxygen (air) directly to electrical energy without relying on the dynamic action of steam heated by reacting fuel-oxygen mixtures, like in steam turbines, or of the reacting gas mixtures themselves, like in gas turbines. The basic rationale for fuel cells is their high efficiencies as compared to indirect-conversion methods. Fuel cells are currently being considered for a number of applications, among them de-centralised power supply. Fuel cells come in five basic types and are usually classified according to the type of electrolyte used, which in turn to a significant degree limits the options for anode and cathode materials. The solid-oxide fuel-cell (SOFC) , with which this thesis is concerned, is thus named after its oxide electrolyte, typically the oxide-ion conducting material yttria-stabilised zirconia (YSZ). While the cathode of an SOFC is often uniform in chemical composition (or at least intended to be), various problems of delamination, cracking etc. associated with the use of metallic anode electrocatalysts led to the development of composite SOFC anodes. Porous anodes consisting of Ni and YSZ particles in roughly 50/50 wt-% mixtures are now almost standard with any SOFC-development programme. The designer of composite SOFC electrodes is faced with at least three, interrelated questions: (1) What will be the optimum microstructure and composition of the composite electrode? (2) If the structure changes during operation, as is often observed, what will be the consequences for the internal losses in the cell? (3) How do we interpret electrochemical and conductivity measurements with regard to structure and composition? It is the primary purpose of this thesis to provide a framework for modelling the electrochemical and transport properties of composite electrodes for SOFC, and to arrive at some new insights that cannot be offered by experiment alone. Emphasis is put on

Computer simulation of variform fuel assemblies using Dragon code

International Nuclear Information System (INIS)

Ju Haitao; Wu Hongchun; Yao Dong

2005-01-01

The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)

Prediction of soot and thermal radiation in a model gas turbine combustor burning kerosene fuel spray at different swirl levels

Science.gov (United States)

Ghose, Prakash; Patra, Jitendra; Datta, Amitava; Mukhopadhyay, Achintya

2016-05-01

Combustion of kerosene fuel spray has been numerically simulated in a laboratory scale combustor geometry to predict soot and the effects of thermal radiation at different swirl levels of primary air flow. The two-phase motion in the combustor is simulated using an Eulerian-Lagragian formulation considering the stochastic separated flow model. The Favre-averaged governing equations are solved for the gas phase with the turbulent quantities simulated by realisable k-ɛ model. The injection of the fuel is considered through a pressure swirl atomiser and the combustion is simulated by a laminar flamelet model with detailed kinetics of kerosene combustion. Soot formation in the flame is predicted using an empirical model with the model parameters adjusted for kerosene fuel. Contributions of gas phase and soot towards thermal radiation have been considered to predict the incident heat flux on the combustor wall and fuel injector. Swirl in the primary flow significantly influences the flow and flame structures in the combustor. The stronger recirculation at high swirl draws more air into the flame region, reduces the flame length and peak flame temperature and also brings the soot laden zone closer to the inlet plane. As a result, the radiative heat flux on the peripheral wall decreases at high swirl and also shifts closer to the inlet plane. However, increased swirl increases the combustor wall temperature due to radial spreading of the flame. The high incident radiative heat flux and the high surface temperature make the fuel injector a critical item in the combustor. The injector peak temperature increases with the increase in swirl flow mainly because the flame is located closer to the inlet plane. On the other hand, a more uniform temperature distribution in the exhaust gas can be attained at the combustor exit at high swirl condition.

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)

Natural circulation in simulated LMFBR fuel assemblies

International Nuclear Information System (INIS)

Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

1985-01-01

Natural circulation experiments have been performed using simulated liquid metal fast breeder reactor fuel assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale sodium loop. Objective of these tests has been to provide experimental data under conditions that might be encountered during a partial or total loss of the shutdown heat removal system (SHRS) in a reactor. The experiments have included single- and two-phase tests under quasi-steady and transient conditions, at both nominal and non-nominal system conditions. Results from these test indicate that the potential for reactor damage during degraded SHRS operation is extremely slight, and that natural circulation can be a major contributor to safe operation of the system in both single- and two-phase flow during such operation

Thermal simulation of a cooling system of hybrid commercial vehicles; Thermalsimulation eine Hybrid-LKW-Kuehlsystems

Energy Technology Data Exchange (ETDEWEB)

Stroh, Christoph; Schnoerch, Stefan; Rathberger, Christian [Magna Powertrain Engineering Center Steyr GmbH und Co. KG, St. Valentin (Austria)

2012-11-01

In the past few years hybrid vehicles have been in the center of automotive engineering efforts, in particular in the field of passenger cars. But hybrid powertrains will also be important for commercial trucks. This focus on hybrid vehicles leads to high demands on thermal management since the additional components in a hybrid vehicle need appropriate cooling or even heating. In the given paper the simulation of a complete cooling system of a hybrid commercial vehicle will be explained. For this virtual examination the commercial 1D thermal management software KULI will be used, a co-simulation with several programs will not be done deliberately. Yet all aspects which are relevant for a global assessment of the thermal management are considered. The main focus is put on the investigation of appropriate concepts for the fluid circuits, including low and high temperature circuits, electric water pumps, etc. Moreover, also a refrigerant circuit with a chiller for active battery cooling will be used, the appropriate control strategy is implemented as well. For simulating transient profiles a simple driving simulation model is included, using road profile, ambient conditions, and various vehicle parameters as input. In addition an engine model is included which enables the investigation of fuel consumption potentials. This simulation model shows how the thermal management of a hybrid vehicle can be investigated with a single program and with reasonable effort. (orig.)

Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

International Nuclear Information System (INIS)

Lima, R. de C.F. de.

1984-01-01

In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author) [pt

A molecular dynamics study of thermal transport in nanoparticle doped Argon like solid

Energy Technology Data Exchange (ETDEWEB)

Shahadat, Muhammad Rubayat Bin, E-mail: rubayat37@gmail.com; Ahmed, Shafkat; Morshed, A. K. M. M. [Department of Mechanical Engineering Bangladesh University of Engineering and Technology (BUET) Dhaka (Bangladesh)

2016-07-12

Interfacial phenomena such as mass and type of the interstitial atom, nano scale material defect influence heat transfer and the effect become very significant with the reduction of the material size. Non Equilibrium Molecular Dynamics (NEMD) simulation was carried out in this study to investigate the effect of the interfacial phenomena on solid. Argon like solid was considered in this study and LJ potential was used for atomic interaction. Nanoparticles of different masses and different molecular defects were inserted inside the solid. From the molecular simulation, it was observed that a large interfacial mismatch due to change in mass in the homogenous solid causes distortion of the phonon frequency causing increase in thermal resistance. Position of the doped nanoparticles have more profound effect on the thermal conductivity of the solid whereas influence of the mass ratio is not very significant. Interstitial atom positioned perpendicular to the heat flow causes sharp reduction in thermal conductivity. Structural defect caused by the molecular defect (void) also observed to significantly affect the thermal conductivity of the solid.

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

Thermal gradient effects on the oxidation of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Klein, A.C.; Reyes, J.N. Jr.; Maguire, M.A.

1990-01-01

A Thermal Gradient Test Facility (TGTF) has been designed and constructed to measure the thermal gradient effect on pressurized water reactor (PWR) fuel rod cladding. The TGTF includes a heat flux simulator assembly capable of producing a wide range of PWR operating conditions including water flow velocities and temperatures, water chemistry conditions, cladding temperatures, and heat fluxes ranging to 160 W/cm 2 . It is fully instrumented including a large number of thermocouples both inside the water flow channel and inside the cladding. Two test programs are in progress. First, cladding specimens are pre-oxidized in air at 500 deg. C and in 400 deg. C steam for various lengths of time to develop a range of uniform oxide thicknesses from 1 to 60 micrometers. The pre-oxidized specimens are placed in the TGTF to characterize the oxide thermal conductivity under a variety of water flow and heat flux conditions. Second, to overcome the long exposure times required under typical PWR conditions a series of tests with the addition of high concentrations of lithium hydroxide to the water are being considered. Static autoclave tests have been conducted with lithium hydroxide concentrations ranging from 0 to 2 moles per liter at 300, 330, and 360 deg. C for up to 36 hours. Results for zircaloy-4 show a considerable increase in the weight gain for the exposed samples with oxidation rate enhancement factors as high as 70 times that of pure water. Operation of the TGTF with elevated lithium hydroxide levels will yield real-time information concerning the effects of a heat flux on the oxidation kinetics of zircaloy fuel rod cladding. (author). 5 refs, 5 figs, 2 tabs

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.

Improvement of numerical simulation methods on safety assessment of the spent fuel storage facility

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Improvement of numerical simulation methods on safety assessment of the spent fuel storage facility is one of main objectives of JNES activities. For the thermal and structural analyses, the radiative heat transfer analysis code S-FOKS has been developed to reduce computing time and to avoid using large memory area. In order to simulate the specular reflection, a new model (called 'model-2') is planned to install to S-FOKS code. The theoretical values with the specular reflection in simple geometry were lead to verify S-FOKS model-2. (author)

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)

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.

Modeling of a thermally integrated 10 kWe planar solid oxide fuel cell system with anode offgas recycling and internal reforming by discretization in flow direction

Science.gov (United States)

Wahl, Stefanie; Segarra, Ana Gallet; Horstmann, Peter; Carré, Maxime; Bessler, Wolfgang G.; Lapicque, François; Friedrich, K. Andreas

2015-04-01

Combined heat and power production (CHP) based on solid oxide fuel cells (SOFC) is a very promising technology to achieve high electrical efficiency to cover power demand by decentralized production. This paper presents a dynamic quasi 2D model of an SOFC system which consists of stack and balance of plant and includes thermal coupling between the single components. The model is implemented in Modelica® and validated with experimental data for the stack UI-characteristic and the thermal behavior. The good agreement between experimental and simulation results demonstrates the validity of the model. Different operating conditions and system configurations are tested, increasing the net electrical efficiency to 57% by implementing an anode offgas recycle rate of 65%. A sensitivity analysis of characteristic values of the system like fuel utilization, oxygen-to-carbon ratio and electrical efficiency for different natural gas compositions is carried out. The result shows that a control strategy adapted to variable natural gas composition and its energy content should be developed in order to optimize the operation of the system.

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)

Numerical Simulation of the Thermal Performance of a Dry Storage Cask for Spent Nuclear Fuel

Directory of Open Access Journals (Sweden)

Heui-Yung Chang

2018-01-01

Full Text Available In this study, the heat flow characteristics and thermal performance of a dry storage cask were investigated via thermal flow experiments and a computational fluid dynamics (CFD simulation. The results indicate that there are many inner circulations in the flow channel of the cask (the channel width is 10 cm. These circulations affect the channel airflow efficiency, which in turn affects the heat dissipation of the dry storage cask. The daily operating temperatures at the top concrete lid and the upper locations of the concrete cask are higher than those permitted by the design specification. The installation of the salt particle collection device has a limited negative effect on the thermal dissipation performance of the dry storage cask.

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

Mechanisms Underpinning Degradation of Protective Oxides and Thermal Barrier Coatings in High Hydrogen Content (HHC) - Fueled Turbines

Energy Technology Data Exchange (ETDEWEB)

Mumm, Daniel

2013-08-31

/thermo-chemical attack mechanisms; (iv) developing a mechanics-based analysis of the driving forces for crack growth and delamination, based on molten phase infiltration, misfit upon cooling, and loss of compliance; (v) understanding changes in TGO growth mechanisms associated with these emerging combustion product streams; and (vi) identifying degradation resistant alternative materials (including new compositions or bi-layer concepts) for use in mitigating the observed degradation modes. To address the materials stability concerns, this program integrated research thrusts aimed at: (1) Conducting tests in simulated syngas and HHC environments to evaluate materials evolution and degradation mechanisms; assessing thermally grown oxide development unique to HHC environmental exposures; carrying out high-resolution imaging and microanalysis to elucidate the evolution of surface deposits (molten phase formation and infiltration); exploring thermo-chemical instabilities; assessing thermo-mechanical drivers and thermal gradient effects on degradation; and quantitatively measuring stress evolution due to enhanced sintering and thermo-chemical instabilities induced in the coating. (2) Executing experiments to study the melting and infiltration of simulated ash deposits, and identifying reaction products and evolving phases associated with molten phase corrosion mechanisms; utilizing thermal spray techniques to fabricate test coupons with controlled microstructures to study mechanisms of instability and degradation; facilitating thermal gradient testing; and developing new materials systems for laboratory testing; (3) Correlating information on the resulting combustion environments to properly assess materials exposure conditions and guide the development of lab-scale simulations of material exposures; specification of representative syngas and high-hydrogen fuels with realistic levels of impurities and contaminants, to explore differences in heat transfer, surface degradation, and deposit

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)

Thermal simulation and validation of 8W LED lamp

NARCIS (Netherlands)

Jakovenko, J.; Werkhoven, R.J.; Formánek, J.; Kunen, J.M.G.; Bolt, P.J.; Kulha, P.

2011-01-01

This work deals with thermal simulation and characterization of solid state lightening (SSL) LED Lamp in order to get precise 3D thermal models for further lamp thermal optimization. Simulations are performed with ANSYS-CFX and CoventorWare software tools. The simulated thermal distribution has been

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.

Development of An Advanced JP-8 Fuel

Science.gov (United States)

1993-12-01

included the Microthermal Precipitation Test (MTP), Fuel Reactor Test, Hot Liquid Process Simulator (HLPS), and Isothermal Corrosion Oxidation Test (ICOT... Microthermal Precipitation Test The impetus for this development effort was the need for a screening test that could discriminate between fuels of...varying propensity to produce thermally induced insoluble particulate material in the bulk fuel. The Microthermal Precipitation (MTP) test thermally

Determining an energy-optimal thermal management strategy for electric driven vehicles

Energy Technology Data Exchange (ETDEWEB)

Suchaneck, Andre; Probst, Tobias; Puente Leon, Fernando [Karlsruher Institut fuer Technology (KIT), Karlsruhe (Germany). Inst. of Industrial Information Technology (IIIT)

2012-11-01

In electric, hybrid electric and fuel cell vehicles, thermal management may have a significant impact on vehicle range. Therefore, optimal thermal management strategies are required. In this paper a method for determining an energy-optimal control strategy for thermal power generation in electric driven vehicles is presented considering all controlled devices (pumps, valves, fans, and the like) as well as influences like ambient temperature, vehicle speed, motor and battery and cooling cycle temperatures. The method is designed to be generic to increase the thermal management development process speed and to achieve the maximal energy reduction for any electric driven vehicle (e.g., by waste heat utilization). Based on simulations of a prototype electric vehicle with an advanced cooling cycle structure, the potential of the method is shown. (orig.)

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

Fully coupled modeling of burnup dependent light water reactor fuel performance using COMSOL Multiphysics

International Nuclear Information System (INIS)

Liu Rong; Zhou Wenzhong; Prudil, Andrew

2015-01-01

This paper presents the development of a light water reactor fuel performance code, which considers almost all the related physical models, 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 equations are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. Comparisons are made for the simulation results between COMSOL and another simulation tool of BISON. The comparisons show the capability of our simulation tool to predict light water UO 2 fuel performances. In our modeling and simulation work, the performance of enhanced thermal conductivity UO 2 -BeO fuel and newly-adopted corrosion resistant SiC cladding material was also studied. UO 2 -BeO high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor would be improved. However, for SiC cladding, although due to its high thermal expansion, the gap closure time is delayed, irradiation induced point defects and defect-clusters in the SiC crystal will dramatically decrease SiC thermal conductivity, and cause significant increase in the fuel temperature. (author)

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

Modeling of the behavior under fuel dispersed irradiation of U-Mo with aluminum matrix from the thermal point of view and its interrelationship with the interdiffusion phase fuel / matrix

International Nuclear Information System (INIS)

Moscarda, Maria V.; Taboada, Horacio H.; Rest, J.

2009-01-01

Results from postirradiation examinations of U-Mo / Al dispersion fuels plates denotes a strong interrelation and feedback between the fuel-matrix interaction and the fuel temperature, bringing undesired consequences on the total swelling and behavior under irradiation. The present work approaches this problem, modeling the profile of temperatures moment by moment to be able to evaluate the increase of this interaction. The Fast Dart program is used, optimized version of program Dart, developed by Dr. J. Rest in collaboration with Dr. H. Taboada. A subroutine of thermal calculation was implemented in this code, which allowed to calculate the evolution of the interaction between the fuel and the matrix. The results of simulations are compared with the results of postirradiation examinations realized by the Reduced Enrichment for Research and Test Reactors International Program. In particular, a good adjustment in the calculation of the depth of interdiffusion U-Mo/Al is observed, demonstrating a right estimation of the profile of temperatures on the fuel plate. It is considered necessary the inclusion of a model that describes the phases that form in the zone of interaction, denoting its thermal dependency and effects due to the radiation damage. (author)

Investigation of typicality of non-nuclear rod and fuel-clad gap effect during reflood phase, and development of a FEM thermal transient analysis code HETFEM

International Nuclear Information System (INIS)

Sudoh, Takashi

1981-06-01

The objective of this study are: 1) Evaluate the capability of the electrical heater for simulating the fuel rod during the reflood phase, and 2) To investigate the effect of the clad-fuel gap in the fuel rod on the clad thermal response during the reflood phase. A computer code HETFEM which is the two dimensional transient thermal conductivity analysis code utilized a finite element method is developed for analysing thermal responses of heater and fuel rod. The two kinds of electrical heaters and a fuel rod are calculated with simple boundary conditions. 1) direct heater (former JAERI reflood test heater), 2) indirect heater (FLECHT test heater), 3) fuel rod (15 x 15 type in Westinghouse PWR). The comparison of the clad temperature responses shows the quench time is influenced by the thermal diffusivity and gap conductance. In the conclusion, the ELECHT heater shows atypicality in the clad temperature response and heat releasing rate. But the direct heater responses are similar to those of the fuel rod. For the gap effect on the fuel rod behavior, the lower gap conductance causes sooner quench and less heat releasing rate. This calculation is not considered the precursory cooling which is affected by heat releasing rate at near and below the quench front. Therefore two dimensional calculation with heat transfer related to the local fluid conditions will be needed. (author)

Lessons Learned From Dynamic Simulations of Advanced Fuel Cycles

International Nuclear Information System (INIS)

Piet, Steven J.; Dixon, Brent W.; Jacobson, Jacob J.; Matthern, Gretchen E.; Shropshire, David E.

2009-01-01

Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe 'lessons learned' from dynamic simulations but attempt to answer the 'so what' question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof

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

Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

International Nuclear Information System (INIS)

Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

1999-01-01

High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

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.

HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

International Nuclear Information System (INIS)

Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

2009-01-01

To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more

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.

Multidimensional simulations of hydrides during fuel rod lifecycle

International Nuclear Information System (INIS)

Stafford, D.S.

2015-01-01

In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim. - Highlights: • We extend BISON fuel performance code to simulate lifecycle of fuel rods. • We model hydrogen evolution in cladding from reactor through dry storage. • We validate 1D simulations of hydrogen evolution against experiments. • We show results of 2D axisymmetric simulations predicting hydride formation. • We show how our model predicts formation of a hydride rim in the cladding.

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}.

International Standard problem ISP 14: behaviour of a fuel bundle simulator during a specified heatup and flooding period (Rebeka experiment): results of post-test analyses: final comparison report

International Nuclear Information System (INIS)

Karwat, H.

1985-02-01

The test consisted in investigating the non-steady material behaviour of a bundle of electrically heated fuel rod simulators with respect to local fuel temperatures, cladding strain, time to burst and local strain at location of burst, together with the thermal hydraulic boundary conditions. The original aim has not been fully achievable. The applied codes for mechanical fuel behaviour largely demonstrated their capabilities for pretest predictions when certain local fluid dynamic parameters are well known to the code users. The difficulties expected with proper analysis of thermal hydraulics of the test were confirmed, caused by the coupling between pin cooling conditions, rod upper plenum calculations and the feedback to clad deformation and burst simulation

Fast-solving thermally thick model of biomass particles embedded in a CFD code for the simulation of fixed-bed burners

International Nuclear Information System (INIS)

Gómez, M.A.; Porteiro, J.; Patiño, D.; Míguez, J.L.

2015-01-01

Highlights: • A thermally thick treatment is used to simulate of fuel the thermal conversion of solid biomass. • A dynamic subgrid scale is used to model the advance of reactive fronts inside the particle. • Efficient solution algorithms are applied to calculate the temperatures and volume of the internal layers. • Several tests were simulated and compared with experimental data. - Abstract: The thermally thick treatment of fuel particles during the thermal conversion of solid biomass is required to consider the internal gradients of temperature and composition and the overlapping of the existing biomass combustion stages. Due to the implied mixture of scales, the balance between model resolution and computational efficiency is an important limitation in the simulation of beds with large numbers of particles. In this study, a subgrid-scale model is applied to consider the intraparticle gradients, the interactions with other particles and the gas phase using a Euler–Euler CFD framework. Numerical heat transfer and mass conservation equations are formulated on a subparticle scale to obtain a system of linear equations that can be used to resolve the temperature and position of the reacting front inside the characteristic particle of each cell. To simulate the entire system, this modelling is combined with other submodels of the gas phase, the bed reaction and the interactions. The performance of the new model is tested using published experimental results for the particle and the bed. Similar temperatures are obtained in the particle-alone tests. Although the mass consumption rates tend to be underpredicted during the drying stage, they are subsequently compensated. In addition, an experimental batch-loaded pellet burner was simulated and tested with different air mass fluxes, in which the experimental ignition rates and temperatures are employed to compare the thermally thick model with the thermally thin model that was previously developed by the authors

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)

A feasibility study on the use of the MOOSE computational framework to simulate three-dimensional deformation of CANDU reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Gamble, Kyle A., E-mail: Kyle.Gamble@inl.gov [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Williams, Anthony F., E-mail: Tony.Williams@cnl.ca [Canadian Nuclear Laboratories, Fuel and Fuel Channel Safety, 1 Plant Road, Chalk River, Ontario, Canada K0J 1J0 (Canada); Chan, Paul K., E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Wowk, Diane, E-mail: Diane.Wowk@rmc.ca [Royal Military College of Canada, Mechanical and Aerospace Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada)

2015-11-15

Highlights: • This is the first demonstration of using the MOOSE framework for modeling CANDU fuel. • Glued and frictionless contact algorithms behave as expected for 2D and 3D cases. • MOOSE accepts and correctly interprets functions of arbitrary form. • 3D deformation calculations accurately compare against analytical solutions. • MOOSE is a viable simulation tool for modeling accident reactor conditions. - Abstract: Horizontally oriented fuel bundles, such as those in CANada Deuterium Uranium (CANDU) reactors present unique modeling challenges. After long irradiation times or during severe transients the fuel elements can laterally deform out of plane due to processes known as bow and sag. Bowing is a thermally driven process that causes the fuel elements to laterally deform when a temperature gradient develops across the diameter of the element. Sagging is a coupled mechanical and thermal process caused by deformation of the fuel pin due to creep mechanisms of the sheathing after long irradiation times and or high temperatures. These out-of-plane deformations can lead to reduced coolant flow and a reduction in coolability of the fuel bundle. In extreme cases element-to-element or element-to-pressure tube contact could occur leading to reduced coolant flow in the subchannels or pressure tube rupture leading to a loss of coolant accident. This paper evaluates the capability of the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at the Idaho National Laboratory to model these deformation mechanisms. The material model capabilities of MOOSE and its ability to simulate contact are also investigated.

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

Optimized Ultrasound-Assisted Oxidative Desulfurization Process of Simulated Fuels over Activated Carbon-Supported Phosphotungstic Acid

Directory of Open Access Journals (Sweden)

Peniel Jean Gildo

2018-01-01

Full Text Available Recent technological advancements respond to the call to minimize/eliminate emissions to the atmosphere. However, on the average, fuel oils which is one of the major raw materials, is found to increase in sulfur concentration due to a phenomenon called thermal maturation. As such, a deeper desulfurization process is needed to obtain low/ultra-low sulfur fuel oils. In the present study, the ultrasound assisted oxidative desulfurization (UAOD processes using the H2O2 and HPW-AC oxidizing system applied to simulated fuel (~2800 ppm sulfur in the form of dibenzothiophene, benzothiophene, and thiophene dissolved in toluene, were optimized. After the pre-saturation of the HPW-AC with the simulated fuel, H2O2 was added just before the reaction was commenced under ultrasonic irradiation. After the application of both 2k-factorial design of experiment for screening and Face-Centered Design of Experiment for optimization, it was found that 25.52 wt% of H2O2 concentration, 983.9 mg of catalyst dose, 9.52 mL aqueous phase per 10 mL of the organic phase and 76.36 minutes of ultrasonication time would render 94.74% oxidation of the sulfur compounds in the simulated fuel. After the application of the optimized parameters to kerosene and employing a 4-cycle extraction using acetonitrile, 99% of the original sulfur content were removed from the kerosene using the UAOD optimized parameters. The desulfurization process resulted in a low-sulfur kerosene which retained its basic fuel properties such as density, viscosity and calorific value.

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

Molecular-Dynamic Simulation In Substation Of Advanced Fuel With Improved Properties

Energy Technology Data Exchange (ETDEWEB)

Kolokol, Alexander S.; Shimkevich, Alexander L. [Russian Research Center ' Kurchatov Institute' , 1 Kurchatov Sq. Moscow 123182 (Russian Federation)

2008-07-01

A disadvantage of the uranium dioxide fuel is very low thermal conductivity than the one of nitride, carbide, metal fuel, and cermets as composites, UO{sub 2}+Me, due to the portion in thermal conductivity of their electronic conductivity and high phonon mobility. An investigation of the microstructure and atomic dynamics of solid solutions as well as the physical and chemical processes in them will make it possible to adjust the properties of the solutions in steps according to prescribed indicators by using alloying additives. The concept for designing an oxide fuel may be promising for the development of a new generation of nuclear reactors. In developing the methods for designing reactor materials as to the nuclear fuel, microscopic structure improving its thermal and physical properties is formulated here. (authors)

Molecular-Dynamic Simulation In Substation Of Advanced Fuel With Improved Properties

International Nuclear Information System (INIS)

Kolokol, Alexander S.; Shimkevich, Alexander L.

2008-01-01

A disadvantage of the uranium dioxide fuel is very low thermal conductivity than the one of nitride, carbide, metal fuel, and cermets as composites, UO 2 +Me, due to the portion in thermal conductivity of their electronic conductivity and high phonon mobility. An investigation of the microstructure and atomic dynamics of solid solutions as well as the physical and chemical processes in them will make it possible to adjust the properties of the solutions in steps according to prescribed indicators by using alloying additives. The concept for designing an oxide fuel may be promising for the development of a new generation of nuclear reactors. In developing the methods for designing reactor materials as to the nuclear fuel, microscopic structure improving its thermal and physical properties is formulated here. (authors)

Simulation study of a PEM fuel cell system fed by hydrogen produced by partial oxidation

Energy Technology Data Exchange (ETDEWEB)

Ozdogan, S [Marmara University, Faculty of Engineering, Istanbul (Turkey); Ersoz, A; Olgun, H [TUBITAK Marmara Research Center, Energy Systems and Environmental Research Institute, Kocaeli (Turkey)

2003-09-01

Within the frame of sustainable development, efficient and clean, if possible zero emission energy production technologies are of utmost importance in various sectors such as utilities, industry, households and transportation. Low-temperature fuel cell systems are suitable for powering transportation systems such as automobiles and trucks in an efficient and low-emitting manner. Proton exchange membrane (PEM) fuel cell systems constitute the most promising low temperature fuel cell option being developed globally. PEM fuel cells generate electric power from air and hydrogen or from a hydrogen rich gas via electrochemical reactions. Water and waste heat are the only by-products of PEM fuel cells. There is great interest in converting current hydrocarbon based common transportation fuels such as gasoline and diesel into hydrogen rich gases acceptable by PEM fuel cells. Hydrogen rich gases can be produced from conventional transportation fuels via various reforming technologies. Steam reforming, partial oxidation and auto-thermal reforming are the three major reforming technologies. In this paper, we discuss the results of a simulation study for a PEM fuel cell with partial oxidation. The Aspen HYSYS 3.1 code has been used for simulation purposes. Two liquid hydrocarbon fuels have been selected to investigate the effect of average molecular weights of hydrocarbons, on the fuel processing efficiency. The overall system efficiency depends on the fuel preparation and fuel cell efficiencies as well as on the heat integration within the system. It is desired to investigate the overall system efficiencies for net electrical power production at 100 kW considering bigger scale transport applications. Results indicate that fuel properties, fuel preparation system operating parameters and PEM fuel cell polarization curve characteristics all affect the overall system efficiency. (authors)

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

Analysis framework to calibrate a numerical model to simulate the thermal test of a 1:2 scale dual purpose cask under accident conditions

International Nuclear Information System (INIS)

Miranda, Carlos A.J.; Libardi, Rosani M.P.; Marcelino, Sergio; Oliveira, Carlos Alberto de; Mattar Neto, Miguel

2013-01-01

This work describes thermal analysis framework including a 3D model and some 2D models to be performed in a 1:2 scale model of a dual-purpose cask to transport and to store spent fuel elements from research reactors to assess the behavior of the cask structure and materials when submitted to heating and drop tests. The analyses should consider all non-linearities involved like the lead phase change and thermal contacts, beside the variation of material properties with the temperature, the air inside it and the heat transfer phenomena (conduction, convection and irradiation) to reproduce the experimental results already obtained in a 1:2 model. A full 3D finite element model takes several hours to run just one analysis. To speed up the analyses to evaluate the significance of some parameters like the emissivity, contact resistance and heat transfer phenomena, among others, two 2D models are planned: one simulating a vertical cut by a diametral plane and another one simulating a horizontal cut by a plane at the cask half height. These 2D models are predicted to run fast enough to allow several analyses in a short period of time and to define options and the best parameters values to match the already obtained experimental results. As this thermal test can not be extrapolated to an 1:1 scale, these parameter values will be used in the final 3D model analysis and also in the full scale model. (author)

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)

Simulation of the effects of grain boundary fission gas during thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Emerson, J.E.; Beiersdorf, B.A.

1984-11-01

This report presents the results of an initial set of out-of-cell transient heating experiments performed on unirradiated UO 2 pellets fabricated to simulate the effect of grain boundary fission gas on fuel swelling and cladding failure. The fabrication involved trapping high-pressure argon on internal pores by sintering annular UO 2 pellets in a hot isostatic press (HIP). The pellet stack was subjected to two separate transients (DGF83-03A and -03B). Figures show photomicrographs of HIPped and non-HIPped UO 2 , respectively, and the adjacent cladding after DGF83-03B. Fuel melting occurred at the center of both the HIPped and non-HIPped pellets; however, a dark ring is present near the center in the HIPped fuel but not in the non-HIPped fuel. This dark band is a high-porosity region due to increased grain boundary/edge swelling in that pellet. In contrast, grain boundary/edge swelling did not occur in the non-HIPped pellets. Thus, the presence of the high-pressure argon trapped on internal pores during sintering in the HIP altered the microstructural behavior. Results of these preliminary tests indicate that the microstructural behavior of HIPped fuel during thermal transients is different from the behavior of conventionally fabricated fuel

Semiquantum molecular dynamics simulation of thermal properties and heat transport in low-dimensional nanostructures

Science.gov (United States)

Savin, Alexander V.; Kosevich, Yuriy A.; Cantarero, Andres

2012-08-01

We present a detailed description of semiquantum molecular dynamics simulation of stochastic dynamics of a system of interacting particles. Within this approach, the dynamics of the system is described with the use of classical Newtonian equations of motion in which the effects of phonon quantum statistics are introduced through random Langevin-like forces with a specific power spectral density (the color noise). The color noise describes the interaction of the molecular system with the thermostat. We apply this technique to the simulation of thermal properties and heat transport in different low-dimensional nanostructures. We describe the determination of temperature in quantum lattice systems, to which the equipartition limit is not applied. We show that one can determine the temperature of such a system from the measured power spectrum and temperature- and relaxation-rate-independent density of vibrational (phonon) states. We simulate the specific heat and heat transport in carbon nanotubes, as well as the heat transport in molecular nanoribbons with perfect (atomically smooth) and rough (porous) edges, and in nanoribbons with strongly anharmonic periodic interatomic potentials. We show that the effects of quantum statistics of phonons are essential for the carbon nanotube in the whole temperature range T<500K, in which the values of the specific heat and thermal conductivity of the nanotube are considerably less than that obtained within the description based on classical statistics of phonons. This conclusion is also applicable to other carbon-based materials and systems with high Debye temperature like graphene, graphene nanoribbons, fullerene, diamond, diamond nanowires, etc. We show that the existence of rough edges and quantum statistics of phonons change drastically the low-temperature thermal conductivity of the nanoribbon in comparison with that of the nanoribbon with perfect edges and classical phonon dynamics and statistics. The semiquantum molecular

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

The 3D thermal-hydraulic numerical simulation for the fuel zone outlet of China experimental fast reactor

International Nuclear Information System (INIS)

Xue Xiuli; Yang Hongyi; Yang Fuchang

2008-01-01

Detailed 3D thermal-hydraulic numerical analyses to the fuel zone outlet are actualized with the STAR-CD CFD code. The performance of sodium mixing is studied and detailed velocity and temperature distribution are obtained in this region which will offer foundations and references to study the rationality of temperature monitoring-spot arrangement and to assess the effect of temperature fluctuations to control rod guide tubes in this region, and so on. (authors)

Simulation and Test of a Fuel Cell Hybrid Golf Cart

Directory of Open Access Journals (Sweden)

Jingming Liang

2014-01-01

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

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

In-core fuel disruption experiments simulating LOF accidents for homogeneous and heterogeneous core LMFBRs: FD2/4 series

International Nuclear Information System (INIS)

Wright, S.A.; Mast, P.K.; Schumacher, Gustav; Fischer, E.A.

1982-01-01

A series of Fuel Disruption (FD) experiments simulating LOF accidents transients for homogeneous- and heterogeneous-core LMFBRs is currently being performed in the Annular Core Research Reactor at SNL. The test fuel is observed with high-speed cinematography to determine the timing and the mode of the fuel disruption. The five experiments performed to date show that the timing and mode of fuel disruption depend on the power level, fuel temperature (after preheat and at disruption), and the fuel temperature gradient. Two basic modes of fuel disruption were observed; solid-state disruption and liquid-state swelling followed by slumping. Solid-state dispersive fuel behavior (several hundred degrees prior to fuel melting) is only observed at high power levels (6P 0 ), low preheat temperatures (2000 K), and high thermal gradients (2800 K/mm). The swelling/slumping behavior was observed in all cases near the time of fuel melting. Computational models have been developed that predict the fuel disruption modes and timing observed in the experiments

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.

Simulating forest fuel and fire risk dynamics across landscapes--LANDIS fuel module design

Science.gov (United States)

Hong S. He; Bo Z. Shang; Thomas R. Crow; Eric J. Gustafson; Stephen R. Shifley

2004-01-01

Understanding fuel dynamics over large spatial (103-106 ha) and temporal scales (101-103 years) is important in comprehensive wildfire management. We present a modeling approach to simulate fuel and fire risk dynamics as well as impacts of alternative fuel treatments. The...

RADYN Simulations of Non-thermal and Thermal Models of Ellerman Bombs

Energy Technology Data Exchange (ETDEWEB)

Hong, Jie; Ding, M. D. [School of Astronomy and Space Science, Nanjing University, Nanjing 210023 (China); Carlsson, Mats, E-mail: dmd@nju.edu.cn [Institute of Theoretical Astrophysics, University of Oslo, P.O. Box 1029 Blindern, NO-0315 Oslo (Norway)

2017-08-20

Ellerman bombs (EBs) are brightenings in the H α line wings that are believed to be caused by magnetic reconnection in the lower atmosphere. To study the response and evolution of the chromospheric line profiles, we perform radiative hydrodynamic simulations of EBs using both non-thermal and thermal models. Overall, these models can generate line profiles that are similar to observations. However, in non-thermal models we find dimming in the H α line wings and continuum when the heating begins, while for the thermal models dimming occurs only in the H α line core, and with a longer lifetime. This difference in line profiles can be used to determine whether an EB is dominated by non-thermal heating or thermal heating. In our simulations, if a higher heating rate is applied, then the H α line will be unrealistically strong and there are still no clear UV burst signatures.

RADYN Simulations of Non-thermal and Thermal Models of Ellerman Bombs

Science.gov (United States)

Hong, Jie; Carlsson, Mats; Ding, M. D.

2017-08-01

Ellerman bombs (EBs) are brightenings in the Hα line wings that are believed to be caused by magnetic reconnection in the lower atmosphere. To study the response and evolution of the chromospheric line profiles, we perform radiative hydrodynamic simulations of EBs using both non-thermal and thermal models. Overall, these models can generate line profiles that are similar to observations. However, in non-thermal models we find dimming in the Hα line wings and continuum when the heating begins, while for the thermal models dimming occurs only in the Hα line core, and with a longer lifetime. This difference in line profiles can be used to determine whether an EB is dominated by non-thermal heating or thermal heating. In our simulations, if a higher heating rate is applied, then the Hα line will be unrealistically strong and there are still no clear UV burst signatures.

Critical heat flux detection in rods simulating fuel elements by using dilation method

International Nuclear Information System (INIS)

Mesquita, A.Z.

1993-01-01

In out-reactor heat transfer experiments, fuel elements are often simulated by electrically heated rods. In order to prevent the heating rod from being damaged by burnout, when the critical heat flux occurs a safety system is provided which checks the axial thermal expansion of the rod. In case of sudden temperature increase, the corresponding elongation causes a fast interruption of the electrical power supply. The experiments presented here show that this method is more effective than one that uses thermocouples. (author)

Detailed Balance of Thermalization Dynamics in Rydberg-Atom Quantum Simulators.

Science.gov (United States)

Kim, Hyosub; Park, YeJe; Kim, Kyungtae; Sim, H-S; Ahn, Jaewook

2018-05-04

Dynamics of large complex systems, such as relaxation towards equilibrium in classical statistical mechanics, often obeys a master equation that captures essential information from the complexities. Here, we find that thermalization of an isolated many-body quantum state can be described by a master equation. We observe sudden quench dynamics of quantum Ising-like models implemented in our quantum simulator, defect-free single-atom tweezers in conjunction with Rydberg-atom interaction. Saturation of their local observables, a thermalization signature, obeys a master equation experimentally constructed by monitoring the occupation probabilities of prequench states and imposing the principle of the detailed balance. Our experiment agrees with theories and demonstrates the detailed balance in a thermalization dynamics that does not require coupling to baths or postulated randomness.

Detailed Balance of Thermalization Dynamics in Rydberg-Atom Quantum Simulators

Science.gov (United States)

Kim, Hyosub; Park, YeJe; Kim, Kyungtae; Sim, H.-S.; Ahn, Jaewook

2018-05-01

Dynamics of large complex systems, such as relaxation towards equilibrium in classical statistical mechanics, often obeys a master equation that captures essential information from the complexities. Here, we find that thermalization of an isolated many-body quantum state can be described by a master equation. We observe sudden quench dynamics of quantum Ising-like models implemented in our quantum simulator, defect-free single-atom tweezers in conjunction with Rydberg-atom interaction. Saturation of their local observables, a thermalization signature, obeys a master equation experimentally constructed by monitoring the occupation probabilities of prequench states and imposing the principle of the detailed balance. Our experiment agrees with theories and demonstrates the detailed balance in a thermalization dynamics that does not require coupling to baths or postulated randomness.

Numerical simulator of the CANDU fueling machine driving desk

International Nuclear Information System (INIS)

Doca, Cezar

2008-01-01

As a national and European premiere, in the 2003 - 2005 period, at the Institute for Nuclear Research Pitesti two CANDU fueling machine heads, no.4 and no.5, for the Nuclear Power Plant Cernavoda - Unit 2 were successfully tested. To perform the tests of these machines, a special CANDU fueling machine testing rig was built and was (and is) available for this goal. The design of the CANDU fueling machine test rig from the Institute for Nuclear Research Pitesti is a replica of the similar equipment operating in CANDU 6 type nuclear power plants. High technical level of the CANDU fueling machine tests required the using of an efficient data acquisition and processing Computer Control System. The challenging goal was to build a computer system (hardware and software) designed and engineered to control the test and calibration process of these fuel handling machines. The design takes care both of the functionality required to correctly control the CANDU fueling machine and of the additional functionality required to assist the testing process. Both the fueling machine testing rig and staff had successfully assessed by the AECL representatives during two missions. At same the time, at the Institute for Nuclear Research Pitesti was/is developed a numerical simulator for the CANDU fueling machine operators training. The paper presents the numerical simulator - a special PC program (software) which simulates the graphics and the functions and the operations at the main desk of the computer control system. The simulator permits 'to drive' a CANDU fueling machine in two manners: manual or automatic. The numerical simulator is dedicated to the training of operators who operate the CANDU fueling machine in a nuclear power plant with CANDU reactor. (author)

Simulation and experimental study on thermal optimization of the heat exchanger for automotive exhaust-based thermoelectric generators

Directory of Open Access Journals (Sweden)

C.Q. Su

2014-11-01

Full Text Available Thermoelectric technology has revealed the potential for automotive exhaust-based thermoelectric generator (TEG, which contributes to the improvement of the fuel economy of the engine-powered vehicle. As a major factor, thermal capacity and heat transfer of the heat exchanger affect the performance of TEG effectively. With the thermal energy of exhaust gas harvested by thermoelectric modules, a temperature gradient appears on the heat exchanger surface, so as the interior flow distribution of the heat exchanger. In order to achieve uniform temperature distribution and higher interface temperature, the thermal characteristics of heat exchangers with various heat transfer enhancement features are studied, such as internal structure, material and surface area. Combining the computational fluid dynamics simulations and infrared test on a high-performance engine with a dynamometer, the thermal performance of the heat exchanger is evaluated. Simulation and experiment results show that a plate-shaped heat exchanger made of brass with accordion-shaped internal structure achieves a relatively ideal performance, which can practically improve overall thermal performance of the TEG.

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.