CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pytel, K.; Mieleszczenko, W.; Lechniak, J.

2010-03-01

The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

DIANA: A multi-phase, multi-component hydrodynamic model for the analysis of severe accidents in heavy water reactors with multiple-tube assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tentner, A.M.

1994-03-01

A detailed hydrodynamic fuel relocation model has been developed for the analysis of severe accidents in Heavy Water Reactors with multiple-tube Assemblies. This model describes the Fuel Disruption and Relocation inside a nuclear fuel assembly and is designated by the acronym DIANA. DIANA solves the transient hydrodynamic equations for all the moving materials in the core and treats all the relevant flow regimes. The numerical solution techniques and some of the physical models included in DIANA have been developed taking advantage of the extensive experience accumulated in the development and validation of the LEVITATE (1) fuel relocation model of SAS4Amore » [2, 3]. The model is designed to handle the fuel and cladding relocation in both voided and partially voided channels. It is able to treat a wide range of thermal/ hydraulic/neutronic conditions and the presence of various flow regimes at different axial locations within the same hydrodynamic channel.« less

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

PWR and BWR spent fuel assembly gamma spectra measurements

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

2016-10-01

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

PWR and BWR spent fuel assembly gamma spectra measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

PWR and BWR spent fuel assembly gamma spectra measurements

DOE PAGES

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

2016-07-17

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Space Shuttle Main Engine structural analysis and data reduction/evaluation. Volume 3B: High pressure fuel turbo-pump preburner pump bearing assembly analysis

NASA Technical Reports Server (NTRS)

Power, Gloria B.; Violett, Rebeca S.

1989-01-01

The analysis performed on the High Pressure Oxidizer Turbopump (HPOTP) preburner pump bearing assembly located on the Space Shuttle Main Engine (SSME) is summarized. An ANSYS finite element model for the inlet assembly was built and executed. Thermal and static analyses were performed.

Creating NDA working standards through high-fidelity spent fuel modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E; Gauld, Ian C; Romano, Catherine E

2012-01-01

The Next Generation Safeguards Initiative (NGSI) is developing advanced non-destructive assay (NDA) techniques for spent nuclear fuel assemblies to advance the state-of-the-art in safeguards measurements. These measurements aim beyond the capabilities of existing methods to include the evaluation of plutonium and fissile material inventory, independent of operator declarations. Testing and evaluation of advanced NDA performance will require reference assemblies with well-characterized compositions to serve as working standards against which the NDA methods can be benchmarked and for uncertainty quantification. To support the development of standards for the NGSI spent fuel NDA project, high-fidelity modeling of irradiated fuel assemblies is beingmore » performed to characterize fuel compositions and radiation emission data. The assembly depletion simulations apply detailed operating history information and core simulation data as it is available to perform high fidelity axial and pin-by-pin fuel characterization for more than 1600 nuclides. The resulting pin-by-pin isotopic inventories are used to optimize the NDA measurements and provide information necessary to unfold and interpret the measurement data, e.g., passive gamma emitters, neutron emitters, neutron absorbers, and fissile content. A key requirement of this study is the analysis of uncertainties associated with the calculated compositions and signatures for the standard assemblies; uncertainties introduced by the calculation methods, nuclear data, and operating information. An integral part of this assessment involves the application of experimental data from destructive radiochemical assay to assess the uncertainty and bias in computed inventories, the impact of parameters such as assembly burnup gradients and burnable poisons, and the influence of neighboring assemblies on periphery rods. This paper will present the results of high fidelity assembly depletion modeling and uncertainty analysis from independent calculations performed using SCALE and MCNP. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.« less

Nuclear imaging of the fuel assembly in ignition experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grim, G. P.; Guler, N.; Merrill, F. E.

First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate thatmore » the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models’ prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.« less

CASMO5/TSUNAMI-3D spent nuclear fuel reactivity uncertainty analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferrer, R.; Rhodes, J.; Smith, K.

2012-07-01

The CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequence in ORNL's SCALE 6 code system to estimate the uncertainties in hot-to-cold reactivity changes due to cross-section uncertainty for PWR assemblies at various burnup points. The goal of the analysis is to establish the multiplication factor uncertainty similarity between various fuel assemblies at different conditions in a quantifiable manner and to obtain a bound on the hot-to-cold reactivity uncertainty over the various assembly types and burnup attributed to fundamental cross-section data uncertainty. (authors)

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

A User’s Guide to the PLTEMP/ANL Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, A. P.; Kalimullah, M.; Feldman, E. E.

2016-07-25

PLTEMP/ANL V4.2 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each withmore » its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.« less

Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

NASA Astrophysics Data System (ADS)

Burdo, James S.

This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the difference was less than 1.00, allowing for the existence of the difference within the margin of error. The second was whether the difference between the two values was big enough to prevent their error bars from overlapping. Error analysis was performed both using a one second count and pseudo-Maxwell statistics for a projected 60 second count, giving four criteria for detection. The number of guide tubes meeting these criteria was compared and graphed for each case. Further analysis at extremes of high and low enrichment and long and short burnup times was done using data from assemblies at the Beaver Valley 1 and 2 PWR. In all neutron flux cases, at least two guide tube locations meet all the criteria for detection of pin diversion. At least one location in almost all of the gamma flux cases does. These results show that placing detectors in the empty guide tubes of spent fuel bundles to identify possible pin diversion is feasible.

Fuel assembly shaker and truck test simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.

2014-09-30

This study continues the modeling support of the SNL shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revisedmore » model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero. The truck test load case had an even lower magnitude than the shaker table case. Strain gage data from the test was compared directly to locations on the model. Truck test strains were lower than the shaker table case, but the model achieved a better relative agreement of 100-200 microstrains (or 0.0001-0.0002 mm/mm). The truck test data included a number of accelerometers at various locations on the truck bed, surrogate basket, and surrogate fuel assembly. This set of accelerometers allowed an evaluation of the dynamics of the conveyance system used in testing. It was discovered that the dynamic load transference through the conveyance has a strong frequency-range dependency. This suggests that different conveyance configurations could behave differently and transmit different magnitudes of loads to the fuel even when traveling down the same road at the same speed. It is recommended that the SNL conveyance system used in testing be characterized through modal analysis and frequency response analysis to provide context and assist in the interpretation of the strain data that was collected during the truck test campaign.« less

Mixed Oxide Fresh Fuel Package Auxiliary Equipment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yapuncich, F.; Ross, A.; Clark, R.H.

2008-07-01

The United States Department of Energy's National Nuclear Security Administration (NNSA) is overseeing the construction the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) on the Savannah River Site. The new facility, being constructed by NNSA's contractor Shaw AREVA MOX Services, will fabricate fuel assemblies utilizing surplus plutonium as feedstock. The fuel will be used in designated commercial nuclear reactors. The MOX Fresh Fuel Package (MFFP), which has recently been licensed by the Nuclear Regulatory Commission (NRC) as a type B package (USA/9295/B(U)F-96), will be utilized to transport the fabricated fuel assemblies from the MFFF to the nuclear reactors. It wasmore » necessary to develop auxiliary equipment that would be able to efficiently handle the high precision fuel assemblies. Also, the physical constraints of the MFFF and the nuclear power plants require that the equipment be capable of loading and unloading the fuel assemblies both vertically and horizontally. The ability to reconfigure the load/unload evolution builds in a large degree of flexibility for the MFFP for the handling of many types of both fuel and non fuel payloads. The design and analysis met various technical specifications including dynamic and static seismic criteria. The fabrication was completed by three major fabrication facilities within the United States. The testing was conducted by Sandia National Laboratories. The unique design specifications and successful testing sequences will be discussed. (authors)« less

Optical fuel pin scanner. [Patent application; for reading identifications

DOEpatents

Kirchner, T.L.; Powers, H.G.

1980-12-09

This patent relates to an optical identification system developed for post-irradiation disassembly and analysis of fuel bundle assemblies. The apparatus is designed to be lowered onto a stationary fuel pin to read identification numbers or letters imprinted on the circumference of the top fuel pin and cap. (DLC)

Outcomes of the JNT 1955 Phase I Viability Study of Gamma Emission Tomography for Spent Fuel Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacobsson-Svard, Staffan; Smith, Leon E.; White, Timothy

The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly has been assessed within the IAEA Support Program project JNT 1955, phase I, which was completed and reported to the IAEA in October 2016. Two safeguards verification objectives were identified in the project; (1) independent determination of the number of active pins that are present in a measured assembly, in the absence of a priori information about the assembly; and (2) quantitative assessment of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup,more » under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives was evaluated across a range of fuel types, burnups and cooling times, while targeting a total interrogation time of less than 60 minutes. The evaluations were founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. The simulated instrument response data were then processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics were defined and used to evaluate the performance of the methods.This paper describes the analysis framework and metrics used to predict tomographer performance. It also presents the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA. Finally, it gives examples of the expected partial-defect detection capabilities for some fuels and diversion scenarios, and it provides a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.« less

Method for monitoring irradiated fuel using Cerenkov radiation

DOEpatents

Dowdy, E.J.; Nicholson, N.; Caldwell, J.T.

1980-05-21

A method is provided for monitoring irradiated nuclear fuel inventories located in a water-filled storage pond wherein the intensity of the Cerenkov radiation emitted from the water in the vicinity of the nuclear fuel is measured. This intensity is then compared with the expected intensity for nuclear fuel having a corresponding degree of irradiation exposure and time period after removal from a reactor core. Where the nuclear fuel inventory is located in an assembly having fuel pins or rods with intervening voids, the Cerenkov light intensity measurement is taken at selected bright sports corresponding to the water-filled interstices of the assembly in the water storage, the water-filled interstices acting as Cerenkov light channels so as to reduce cross-talk. On-line digital analysis of an analog video signal is possible, or video tapes may be used for later measurement using a video editor and an electrometer. Direct measurement of the Cerenkov radiation intensity also is possible using spot photometers pointed at the assembly.

Fuel rod assembly to manifold attachment

DOEpatents

Donck, Harry A.; Veca, Anthony R.; Snyder, Jr., Harold J.

1980-01-01

A fuel element is formed with a plurality of fuel rod assemblies detachably connected to an overhead support with each of the fuel rod assemblies having a gas tight seal with the support to allow internal fission gaseous products to flow without leakage from the fuel rod assemblies into a vent manifold passageway system on the support. The upper ends of the fuel rod assemblies are located at vertically extending openings in the support and upper threaded members are threaded to the fuel rod assemblies to connect the latter to the support. The preferred threaded members are cap nuts having a dome wall encircling an upper threaded end on the fuel rod assembly and having an upper sealing surface for sealing contact with the support. Another and lower seal is achieved by abutting a sealing surface on each fuel rod assembly with the support. A deformable portion on the cap nut locks the latter against inadvertent turning off the fuel rod assembly. Orienting means on the fuel rod and support primarily locates the fuel rods azimuthally for reception of a deforming tool for the cap nut. A cross port in the fuel rod end plug discharges into a sealed annulus within the support, which serves as a circumferential chamber, connecting the manifold gas passageways in the support.

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE PAGES

Ilas, Germina; Liljenfeldt, Henrik

2017-05-19

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Liljenfeldt, Henrik

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Fuel injection assembly for use in turbine engines and method of assembling same

DOEpatents

Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

2015-12-15

A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

Control assembly for controlling a fuel cell system during shutdown and restart

DOEpatents

Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

2010-06-15

A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

Fuel injection assembly for gas turbine engine combustor

NASA Technical Reports Server (NTRS)

Candy, Anthony J. (Inventor); Glynn, Christopher C. (Inventor); Barrett, John E. (Inventor)

2002-01-01

A fuel injection assembly for a gas turbine engine combustor, including at least one fuel stem, a plurality of concentrically disposed tubes positioned within each fuel stem, wherein a cooling supply flow passage, a cooling return flow passage, and a tip fuel flow passage are defined thereby, and at least one fuel tip assembly connected to each fuel stem so as to be in flow communication with the flow passages, wherein an active cooling circuit for each fuel stem and fuel tip assembly is maintained by providing all active fuel through the cooling supply flow passage and the cooling return flow passage during each stage of combustor operation. The fuel flowing through the active cooling circuit is then collected so that a predetermined portion thereof is provided to the tip fuel flow passage for injection by the fuel tip assembly.

Analysis of key safety metrics of thorium utilization in LWRs

DOE PAGES

Ade, Brian J.; Bowman, Stephen M.; Worrall, Andrew; ...

2016-04-08

Here, thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel ( 233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based onmore » concepts that mix thorium with uranium (UO 2 + ThO 2) or that add fertile thorium (ThO 2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO 2 + ThO 2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO 2, alone or mixed with UO 2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO 2 + UO 2 and ThO 2 + PuO 2 against those of UO 2 and typical UO 2 + PuO 2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO 2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of knowledge gaps that require additional analysis or research to develop a technical basis for the licensing of thorium fuel are identified.« less

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

2016-12-01

The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilles Youinou; Andrea Alfonsi

2012-03-01

This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis,more » the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.« less

CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toth, Sandor; Legradi, Gabor; Aszodi, Attila

2006-07-01

From the aspect of planning the power upgrading of nuclear reactors - including the VVER-440 type reactor - it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960more » mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper. (authors)« less

76 FR 12825 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Confirmation of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-09

... definitions for Damaged Fuel Assembly and Transfer Operations; add definitions for Fuel Class and Reconstituted Fuel Assembly; add Combustion Engineering 16x16 class fuel assemblies as authorized contents...

Locking support for nuclear fuel assemblies

DOEpatents

Ledin, Eric

1980-01-01

A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

77 FR 70114 - Airworthiness Directives; Cessna Aircraft Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-23

... assemblies, which were caused by the fuel return line assembly rubbing against the right steering tube assembly during full rudder pedal actuation. This AD requires you to inspect the fuel return line assembly... the fuel return line assembly and both the right steering tube assembly and the airplane structure...

ADVANCEMENTS IN TIME-SPECTRA ANALYSIS METHODS FOR LEAD SLOWING-DOWN SPECTROSCOPY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Leon E.; Anderson, Kevin K.; Gesh, Christopher J.

2010-08-11

Direct measurement of Pu in spent nuclear fuel remains a key challenge for safeguarding nuclear fuel cycles of today and tomorrow. Lead slowing-down spectroscopy (LSDS) is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic mass with an uncertainty lower than the approximately 10 percent typical of today’s confirmatory assay methods. Pacific Northwest National Laboratory’s (PNNL) previous work to assess the viability of LSDS for the assay of pressurized water reactor (PWR) assemblies indicated that the method could provide direct assay of Pu-239 and U-235 (and possibly Pu-240 and Pu-241)more » with uncertainties less than a few percent, assuming suitably efficient instrumentation, an intense pulsed neutron source, and improvements in the time-spectra analysis methods used to extract isotopic information from a complex LSDS signal. This previous simulation-based evaluation used relatively simple PWR fuel assembly definitions (e.g. constant burnup across the assembly) and a constant initial enrichment and cooling time. The time-spectra analysis method was founded on a preliminary analytical model of self-shielding intended to correct for assay-signal nonlinearities introduced by attenuation of the interrogating neutron flux within the assembly.« less

Method for shearing spent nuclear fuel assemblies

DOEpatents

Weil, Bradley S.; Watson, Clyde D.

1977-01-01

A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Root, M. A.; Menlove, H. O.; Lanza, R. C.

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE PAGES

Root, M. A.; Menlove, H. O.; Lanza, R. C.; ...

2018-03-21

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lubina, A. S., E-mail: lubina-as@nrcki.ru; Subbotin, A. S.; Sedov, A. A.

2016-12-15

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The resultsmore » of the hydrodynamics and heat transfer calculations have been analyzed.« less

Fuel cell sub-assembly

DOEpatents

Chi, Chang V.

1983-01-01

A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

DOEpatents

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

Improved nuclear fuel assembly grid spacer

DOEpatents

Marshall, John; Kaplan, Samuel

1977-01-01

An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

Application of X-Ray Computer Tomography for Observing the Central Void Formations and the Fuel Pin Deformations of Irradiated FBR Fuel Assemblies

NASA Astrophysics Data System (ADS)

Katsuyama, Kozo; Nagamine, Tsuyoshi; Furuya, Hirotaka

2010-10-01

In order to observe the structural change in the interior of irradiated fuel assemblies, a non-destructive post-irradiation examination (PIE) technique using X-ray computer tomography (X-ray CT) was developed. This X-ray CT technique was applied to observe the central void formations and fuel pin deformations of fuel assemblies which had been irradiated at high linear heat rating. The central void sizes in all fuel pins were measured on five cross sections of the core fuel column as a parameter for evaluating fuel thermal performance. In addition, the fuel pin deformations were analyzed from X-ray CT images obtained along the axial direction of a fuel assembly at the same separation interval. A dependence of void size on the linear heat rating was seen in the fuel assembly irradiated at high linear heat rating. In addition, significant undulations of the fuel pin were observed along the axial direction, coinciding with the wrapping wire pitch in the core fuel column. Application of the developed technique should provide enhanced resolution of measurements and simplify fuel PIEs.

ANALYSIS AND EXAMINATION OF MOX FUEL FROM NONPROLIFERATION PROGRAMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCoy, Kevin; Machut, Dr McLean; Morris, Robert Noel

The U.S. Department of Energy has decided to dispose of a portion of the nation s surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. Four lead assemblies were manufactured and irradiated to a maximum fuel rod burnup of 47.3 MWd/kg heavy metal. This was the first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio of less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. The performance of the rodsmore » was analyzed with AREVA s next-generation GALILEO code. The results of the analysis confirmed that the fuel rods had performed safely and predictably, and that GALILEO is applicable to MOX fuel with a low 240Pu/239Pu ratio as well as to standard MOX. The results are presented and compared to the GALILEO database. In addition, the fuel cladding was tested to confirm that traces of gallium in the fuel pellets had not affected the mechanical properties of the cladding. The irradiated cladding was found to remain ductile at both room temperature and 350 C for both the axial and circumferential directions.« less

Molybdenum-99 production calculation analysis of SAMOP reactor based on thorium nitrate fuel

NASA Astrophysics Data System (ADS)

Syarip; Togatorop, E.; Yassar

2018-03-01

SAMOP (Subcritical Assembly for Molybdenum-99 Production) has the potential to use thorium as fuel to produce 99Mo after modifying the design, but the production performance has not been discovered yet. A study needs to be done to obtain the correlation between 99Mo production with the mixed fuel composition of uranium and with SAMOP power on the modified SAMOP design. The study aims to obtain the production of 99Mo based thorium nitrate fuel on SAMOP’s modified designs. Monte Carlo N-Particle eXtended (MCNPX) is required to simulate the operation of the assembly by varying the composition of the uranium-thorium nitrate mixed fuel, geometry and power fraction on the SAMOP modified designs. The burnup command on the MCNPX is used to confirm the 99Mo production result. The assembly is simulated to operate for 6 days with subcritical neutron multiplication factor (keff = 0.97-0.99). The neutron multiplication factor of the modified design (keff) is 0.97, the activity obtained from 99Mo is 18.58 Ci at 1 kW power operation.

In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; De Baere, Paul

2015-04-17

A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hourmore » of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.« less

System and method for controlling a combustor assembly

DOEpatents

York, William David; Ziminsky, Willy Steve; Johnson, Thomas Edward; Stevenson, Christian Xavier

2013-03-05

A system and method for controlling a combustor assembly are disclosed. The system includes a combustor assembly. The combustor assembly includes a combustor and a fuel nozzle assembly. The combustor includes a casing. The fuel nozzle assembly is positioned at least partially within the casing and includes a fuel nozzle. The fuel nozzle assembly further defines a head end. The system further includes a viewing device configured for capturing an image of at least a portion of the head end, and a processor communicatively coupled to the viewing device, the processor configured to compare the image to a standard image for the head end.

Swelling-resistant nuclear fuel

DOEpatents

Arsenlis, Athanasios [Hayward, CA; Satcher, Jr., Joe; Kucheyev, Sergei O [Oakland, CA

2011-12-27

A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

40 CFR 1060.102 - What permeation emission control requirements apply for fuel lines?

Code of Federal Regulations, 2011 CFR

2011-07-01

... assemblies as aggregated systems that include multiple sections of fuel line with connectors and fittings. For example, you may certify fuel lines for portable marine fuel tanks as assemblies of fuel hose, primer bulbs, and self-sealing end connections. The length of such an assembly must not be longer than a...

40 CFR 1060.102 - What permeation emission control requirements apply for fuel lines?

Code of Federal Regulations, 2010 CFR

2010-07-01

... assemblies as aggregated systems that include multiple sections of fuel line with connectors and fittings. For example, you may certify fuel lines for portable marine fuel tanks as assemblies of fuel hose, primer bulbs, and self-sealing end connections. The length of such an assembly must not be longer than a...

Fabrication of fuel pin assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1972-01-01

Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE PAGES

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

2017-09-01

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pond, R.B.; Matos, J.E.

1996-05-01

As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, aremore » not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.« less

An anisotropic numerical model for thermal hydraulic analyses: application to liquid metal flow in fuel assemblies

NASA Astrophysics Data System (ADS)

Vitillo, F.; Vitale Di Maio, D.; Galati, C.; Caruso, G.

2015-11-01

A CFD analysis has been carried out to study the thermal-hydraulic behavior of liquid metal coolant in a fuel assembly of triangular lattice. In order to obtain fast and accurate results, the isotropic two-equation RANS approach is often used in nuclear engineering applications. A different approach is provided by Non-Linear Eddy Viscosity Models (NLEVM), which try to take into account anisotropic effects by a nonlinear formulation of the Reynolds stress tensor. This approach is very promising, as it results in a very good numerical behavior and in a potentially better fluid flow description than classical isotropic models. An Anisotropic Shear Stress Transport (ASST) model, implemented into a commercial software, has been applied in previous studies, showing very trustful results for a large variety of flows and applications. In the paper, the ASST model has been used to perform an analysis of the fluid flow inside the fuel assembly of the ALFRED lead cooled fast reactor. Then, a comparison between the results of wall-resolved conjugated heat transfer computations and the results of a decoupled analysis using a suitable thermal wall-function previously implemented into the solver has been performed and presented.

LMFBR fuel assembly design for HCDA fuel dispersal

DOEpatents

Lacko, Robert E.; Tilbrook, Roger W.

1984-01-01

A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

Methods of conditioning direct methanol fuel cells

DOEpatents

Rice, Cynthia; Ren, Xiaoming; Gottesfeld, Shimshon

2005-11-08

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

Flashback resistant pre-mixer assembly

DOEpatents

Laster, Walter R [Oviedo, FL; Gambacorta, Domenico [Oviedo, FL

2012-02-14

A pre-mixer assembly associated with a fuel supply system for mixing of air and fuel upstream from a main combustion zone in a gas turbine engine. The pre-mixer assembly includes a swirler assembly disposed about a fuel injector of the fuel supply system and a pre-mixer transition member. The swirler assembly includes a forward end defining an air inlet and an opposed aft end. The pre-mixer transition member has a forward end affixed to the aft end of the swirler assembly and an opposed aft end defining an outlet of the pre-mixer assembly. The aft end of the pre-mixer transition member is spaced from a base plate such that a gap is formed between the aft end of the pre-mixer transition member and the base plate for permitting a flow of purge air therethrough to increase a velocity of the air/fuel mixture exiting the pre-mixer assembly.

DOE Office of Scientific and Technical Information (OSTI.GOV)

West, W.S.

Progress during the period includes completion of the SNAP 7C system tests, completion of safety analysis for the SNAP 7A and C systems, assembly and initial testing of SNAP 7A, assembly of a modified reliability model, and assembly of a 10-W generator. Other activities include completion of thermal and safety analyses for SNAP 7B and D generators and fuel processing for these generators. (J.R.D.)

Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

DOEpatents

Bradley, John G.

1982-01-01

A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

Interface ring for gas turbine fuel nozzle assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, Timothy A.; Schilp, Reinhard

A gas turbine combustor assembly including a combustor liner and a plurality of fuel nozzle assemblies arranged in an annular array extending within the combustor liner. The fuel nozzle assemblies each include fuel nozzle body integral with a swirler assembly, and the swirler assemblies each include a bellmouth structure to turn air radially inwardly for passage into the swirler assemblies. A radially outer removed portion of each of the bellmouth structures defines a periphery diameter spaced from an inner surface of the combustor liner, and an interface ring is provided extending between the combustor liner and the removed portions ofmore » the bellmouth structures at the periphery diameter.« less

MARTINS: A foam/film flow model for molten material relocation in HWRs with U-Al-fueled multi-tube assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalimullah

1994-03-01

Some special purpose heavy-water reactors (EM) are made of assemblies consisting of a number of coaxial aluminum-clad U-Al alloy fuel tubes and an outer Al sleeve surrounding the fuel tubes. The heavy water coolant flows in the annular gaps between the circular tubes. Analysis of severe accidents in such reactors requires a model for predicting the behavior of the fuel tubes as they melt and disrupt. This paper describes a detailed, mechanistic model for fuel tube heatup, melting, freezing, and molten material relocation, called MARTINS (Melting and Relocation of Tubes in Nuclear subassembly). The paper presents the modeling of themore » phenomena in MARTINS, and an application of the model to analysis of a reactivity insertion accident. Some models are being developed to compute gradual downward relocation of molten material at decay-heat power levels via candling along intact tubes, neglecting coolant vapor hydrodynamic forces on molten material. These models are inadequate for high power accident sequences involving significant hydrodynamic forces. These forces are included in MARTINS.« less

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE PAGES

Ham, Y.; Kerr, P.; Sitaraman, S.; ...

2016-05-05

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Y.S.; Kerr, P.; Sitaraman, S.

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported themore » successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)« less

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Y.; Kerr, P.; Sitaraman, S.

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Inactive end cell assembly for fuel cells for improved electrolyte management and electrical contact

DOEpatents

Yuh, Chao-Yi [New Milford, CT; Farooque, Mohammad [Danbury, CT; Johnsen, Richard [New Fairfield, CT

2007-04-10

An assembly for storing electrolyte in a carbonate fuel cell is provided. The combination of a soft, compliant and resilient cathode current collector and an inactive anode part including a foam anode in each assembly mitigates electrical contact loss during operation of the fuel cell stack. In addition, an electrode reservoir in the positive end assembly and an electrode sink in the negative end assembly are provided, by which ribbed and flat cathode members inhibit electrolyte migration in the fuel cell stack.

X-ray source assembly having enhanced output stability, and fluid stream analysis applications thereof

DOEpatents

Radley, Ian [Glenmont, NY; Bievenue, Thomas J [Delmar, NY; Burdett, John H [Charlton, NY; Gallagher, Brian W [Guilderland, NY; Shakshober, Stuart M [Hudson, NY; Chen, Zewu [Schenectady, NY; Moore, Michael D [Alplaus, NY

2008-06-08

An x-ray source assembly and method of operation are provided having enhanced output stability. The assembly includes an anode having a source spot upon which electrons impinge and a control system for controlling position of the anode source spot relative to an output structure. The control system can maintain the anode source spot location relative to the output structure notwithstanding a change in one or more operating conditions of the x-ray source assembly. One aspect of the disclosed invention is most amenable to the analysis of sulfur in petroleum-based fuels.

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

NASA Astrophysics Data System (ADS)

Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

2016-08-01

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel assemblies. Based on analysis of modern engineering solutions on SNF regeneration, it has been concluded that new reprocessing technologies should meet the ecological safety requirements, provide a more extensive use of the resource base of nuclear engineering, allow the production of valuable and trace elements on an industrial scale, and decrease radioactive waste release.

Spent nuclear fuel assembly inspection using neutron computed tomography

NASA Astrophysics Data System (ADS)

Pope, Chad Lee

The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE PAGES

Jiang, Hao; Wang, Jy-An John

2016-10-14

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Investigation of Ruthenium Dissolution in Advanced Membrane Electrode Assemblies for Direct Methanol Based Fuel Cells Stacks

NASA Technical Reports Server (NTRS)

Valdez, T. I.; Firdosy, S.; Koel, B. E.; Narayanan, S. R.

2005-01-01

This viewgraph presentation gives a detailed review of the Direct Methanol Based Fuel Cell (DMFC) stack and investigates the Ruthenium that was found at the exit of the stack. The topics include: 1) Motivation; 2) Pathways for Cell Degradation; 3) Cell Duration Testing; 4) Duration Testing, MEA Analysis; and 5) Stack Degradation Analysis.

System and method for injecting fuel

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2012-12-04

According to various embodiments, a system includes a staggered multi-nozzle assembly. The staggered multi-nozzle assembly includes a first fuel nozzle having a first axis and a first flow path extending to a first downstream end portion, wherein the first fuel nozzle has a first non-circular perimeter at the first downstream end portion. The staggered multi-nozzle assembly also includes a second fuel nozzle having a second axis and a second flow path extending to a second downstream end portion, wherein the first and second downstream end portions are axially offset from one another relative to the first and second axes. The staggered multi-nozzle assembly further includes a cap member disposed circumferentially about at least the first and second fuel nozzles to assemble the staggered multi-nozzle assembly.

Fuel assembly for nuclear reactors

DOEpatents

Creagan, Robert J.; Frisch, Erling

1977-01-01

A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

Spent fuel burnup estimation by Cerenkov glow intensity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki

1994-10-01

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

NASA Astrophysics Data System (ADS)

Widargo, Reza; Anghaie, Samim

1999-01-01

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight ratio.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wiebe, David J; Fox, Timothy A

A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passagemore » is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.« less

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, M.L.; Rosenstein, R.G.

1998-10-13

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

Mox fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-05-15

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-07-17

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

1998-01-01

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

PubMed

Rezaeian, M; Kamali, J

2017-01-01

Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B 4 C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

Reduced size fuel cell for portable applications

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor); Clara, Filiberto (Inventor); Frank, Harvey A. (Inventor)

2004-01-01

A flat pack type fuel cell includes a plurality of membrane electrode assemblies. Each membrane electrode assembly is formed of an anode, an electrolyte, and an cathode with appropriate catalysts thereon. The anode is directly into contact with fuel via a wicking element. The fuel reservoir may extend along the same axis as the membrane electrode assemblies, so that fuel can be applied to each of the anodes. Each of the fuel cell elements is interconnected together to provide the voltage outputs in series.

Coolant mass flow equalizer for nuclear fuel

DOEpatents

Betten, Paul R.

1978-01-01

The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Trahan, Alexis Chanel

New nondestructive assay techniques are sought to better characterize spent nuclear fuel. One of the NDA instruments selected for possible deployment is differential die-away self-interrogation (DDSI). The proposed DDSI approach for spent fuel assembly assay utilizes primarily the spontaneous fission and (alpha, n) neutrons in the assemblies as an internal interrogating radiation source. The neutrons released in spontaneous fission or (alpha,n) reactions are thermalized in the surrounding water and induce fission in fissile isotopes, thereby creating a measurable signal from isotopes of interest that would be otherwise difficult to measure. The DDSI instrument employs neutron coincidence counting with 3He tubes and list-mode-based data acquisition to allow for production of Rossi-alpha distributions (RADs) in post-processing. The list-mode approach to data collection and subsequent construction of RADs has expanded the analytical possibilities, as will be demonstrated throughout this thesis. One of the primary advantages is that the measured signal in the form of a RAD can be analyzed in its entirety including determination of die-away times in different time domains. This capability led to the development of the early die-away method, a novel leakage multiplication determination method which is tested throughout the thesis on different sources in simulation space and fresh fuel experiments. The early die-away method is a robust, accurate, improved method of determining multiplication without the need for knowledge of the (alpha,n) source term. The DDSI technique and instrument are presented along with the many novel capabilities enabled by and discovered through RAD analysis. Among the new capabilities presented are the early die-away method, total plutonium content determination, and highly sensitive missing pin detection. Simulation of hundreds of different spent and fresh fuel assemblies were used to develop the analysis algorithms and the techniques were tested on a variety of spontaneous fission-driven fresh fuel assemblies at Los Alamos National Laboratory and the BeRP ball at the Nevada National Security Site. The development of the new, improved analysis and characterization methods with the DDSI instrument makes it a viable technique for implementation in a facility to meet material control and safeguards needs.

Shock ignition of thermonuclear fuel with high areal density.

PubMed

Betti, R; Zhou, C D; Anderson, K S; Perkins, L J; Theobald, W; Solodov, A A

2007-04-13

A novel method by C. Zhou and R. Betti [Bull. Am. Phys. Soc. 50, 140 (2005)] to assemble and ignite thermonuclear fuel is presented. Massive cryogenic shells are first imploded by direct laser light with a low implosion velocity and on a low adiabat leading to fuel assemblies with large areal densities. The assembled fuel is ignited from a central hot spot heated by the collision of a spherically convergent ignitor shock and the return shock. The resulting fuel assembly features a hot-spot pressure greater than the surrounding dense fuel pressure. Such a nonisobaric assembly requires a lower energy threshold for ignition than the conventional isobaric one. The ignitor shock can be launched by a spike in the laser power or by particle beams. The thermonuclear gain can be significantly larger than in conventional isobaric ignition for equal driver energy.

Method of preparing gas tags for identification of single and multiple failures of nuclear reactor fuel assemblies

DOEpatents

McCormick, Norman J.

1976-01-01

For use in the identification of failed fuel assemblies in a nuclear reactor, the ratios of the tag gas isotopic concentrations are located on curved surfaces to enable the ratios corresponding to failure of a single fuel assembly to be distinguished from those formed from any combination of two or more failed assemblies.

Turbine combustor with fuel nozzles having inner and outer fuel circuits

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward; Kim, Kwanwoo

2013-12-24

A combustor cap assembly for a turbine engine includes a combustor cap and a plurality of fuel nozzles mounted on the combustor cap. One or more of the fuel nozzles would include two separate fuel circuits which are individually controllable. The combustor cap assembly would be controlled so that individual fuel circuits of the fuel nozzles are operated or deliberately shut off to provide for physical separation between the flow of fuel delivered by adjacent fuel nozzles and/or so that adjacent fuel nozzles operate at different pressure differentials. Operating a combustor cap assembly in this fashion helps to reduce or eliminate the generation of undesirable and potentially harmful noise.

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mertyurek, Ugur; Gauld, Ian C.

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE PAGES

Mertyurek, Ugur; Gauld, Ian C.

2015-12-24

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

X-ray source assembly having enhanced output stability, and fluid stream analysis applications thereof

DOEpatents

Radley, Ian; Bievenue, Thomas J.; Burdett Jr., John H.; Gallagher, Brian W.; Shakshober, Stuart M.; Chen, Zewu; Moore, Michael D.

2007-04-24

An x-ray source assembly (2700) and method of operation are provided having enhanced output stability. The assembly includes an anode (2125) having a source spot upon which electrons (2120) impinge and a control system (2715/2720) for controlling position of the anode source spot relative to an output structure. The control system can maintain the anode source spot location relative to the output structure (2710) notwithstanding a change in one or more operating conditions of the x-ray source assembly. One aspect of the disclosed invention is most amenable to the analysis of sulfur in petroleum-based fuels.

Measurement of the 235U Induced Fission Gamma-ray Spectrum as an Active Non-destructive Assay of Fresh Nucleear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarnoski, Sarah E.; Fast, James E.; Fulsom, Bryan G.

2017-07-17

Non-destructive assay is a powerful tool the International Atomic Energy Agency (IAEA) employs to verify adherence to safeguards agreements. Current IAEA veri- cation techniques for fresh nuclear fuel include passive gamma-ray spectroscopy to determine fuel enrichment. This technique suers from self-shielding and lakes the percision to detect diversion of central fuel rods. The aim of this research is to develop a new, more capable non-destructive analysis technique using active neutron interroga- tion of fuel assemblies and determining the yields of short-lived ssion products from high-resolution gamma-ray spectroscopy using high-purity germanium (HPGe). This paper reports results from irradiation of a onemore » meter tall mock fresh fuel assembly with low enriched uranium (LEU) or depleted uranium (DU) rods using a down-scattered deuterium-tritium (D-T) neutron source. Both prompt and delayed gamma-ray spec- tra were collected as time-stamped list-mode data in a coax detector and without list mode data in a planar strip detector. No dierentiating signatures were observed in the prompt spectra in either detector; however, both detectors observed several short-lived ssion product signatures in LEU and not DU fuel, indicating that this technique has potential for determination of enrichment of fresh fuel assemblies. There were eight unique ssion products observed in the LEU spectra with the coax detector spectra, and three ssion products were observed in the LEU spectra with the strip detector.« less

Fuel and oxidizer valve assembly employs single solenoid actuator

NASA Technical Reports Server (NTRS)

1966-01-01

Valve assembly simultaneously starts or stops the flow of oxidizer and fuel from separate inlet channels to reaction control motors. The assembly combines an oxidizer shutoff valve and a fuel shutoff valve which are mechanically linked and operated by a single high-speed solenoid actuator.

Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.

2012-07-01

The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now.more » The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)« less

Modular fuel-cell stack assembly

DOEpatents

Patel, Pinakin

2010-07-13

A fuel cell assembly having a plurality of fuel cells arranged in a stack. An end plate assembly abuts the fuel cell at an end of said stack. The end plate assembly has an inlet area adapted to receive an exhaust gas from the stack, an outlet area and a passage connecting the inlet area and outlet area and adapted to carry the exhaust gas received at the inlet area from the inlet area to the outlet area. A further end plate assembly abuts the fuel cell at a further opposing end of the stack. The further end plate assembly has a further inlet area adapted to receive a further exhaust gas from the stack, a further outlet area and a further passage connecting the further inlet area and further outlet area and adapted to carry the further exhaust gas received at the further inlet area from the further inlet area to the further outlet area.

Method and apparatus for close packing of nuclear fuel assemblies

DOEpatents

Newman, Darrell F.

1993-01-01

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Method and apparatus for close packing of nuclear fuel assemblies

DOEpatents

Newman, D.F.

1993-03-30

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knaab, H.; Knecht, K.

The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

Piloted rich-catalytic lean-burn hybrid combustor

DOEpatents

Newburry, Donald Maurice

2002-01-01

A catalytic combustor assembly which includes, an air source, a fuel delivery means, a catalytic reactor assembly, a mixing chamber, and a means for igniting a fuel/air mixture. The catalytic reactor assembly is in fluid communication with the air source and fuel delivery means and has a fuel/air plenum which is coated with a catalytic material. The fuel/air plenum has cooling air conduits passing therethrough which have an upstream end. The upstream end of the cooling conduits is in fluid communication with the air source but not the fuel delivery means.

Some methods for achieving more efficient performance of fuel assemblies

NASA Astrophysics Data System (ADS)

Boltenko, E. A.

2014-07-01

More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.

Assess How Changes in Fuel Cycle Operation Impact Safeguards

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph; Adigun, Babatunde John; Fugate, Michael Lynn

Since the beginning of commercial nuclear power generation in the 1960s, the ability of researchers to understand and control the isotopic content of spent fuel has improved. It is therefore not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Modern fuelmore » therefore should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and, as such, the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements. The status quo for spent fuel assay in the safeguards context is that the overwhelming majority of spent fuel assemblies are not measured in a quantitative way except for those assemblies about to be loaded into a difficult or impossible to access location (dry storage or, in the future, a repository). In other words, when the assembly is still accessible to a state actor, or an insider, when it is cooling in a pool, the inspectorate does not have a measurement database that could assist them in re-­verifying the integrity of that assembly. The spent fuel safeguards regime would be strengthened if spent fuel assemblies were measured from discharge to loading into a difficult or impossible to access location. The primary driver for suggesting this shift in approach is the change in robotic technology and information technology in general. It should be possible, with minimal impact to the facility, to measure each assembly every time that it is moved in the pool, with the first measurements being made at discharge. The following conclusions were reached: The total neutron count rate can be accurately predicted at any future moment in time based upon the measured count rate at discharge, provided the initial enrichment and burnup of the assembly is known at discharge. It is expected that the total neutron count rate measured at discharge will be indicative of the initial enrichment and burnup of that assembly. If the automated robot were to focus on measuring the assemblies in the rack without moving them, the time available would increase immensely.« less

A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

NASA Astrophysics Data System (ADS)

Bushuev, A. V.; Kozhin, A. F.; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E.

2016-12-01

An active neutron method for measuring the residual mass of 235U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual 235U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of 238U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

NASA Astrophysics Data System (ADS)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

Structural integrity assessment and stress measurement of chasnupp-1 fuel assembly skeleton: under tensile loading condition

NASA Astrophysics Data System (ADS)

Waseem; Siddiqui, Ashfaq Ahmad; Murtaza, Ghulam; Maqbool, Abu Baker

2017-12-01

Fuel assembly (FA) structure without fuel rods is called FA skeleton which is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the Chashma Nuclear power plant-1 FA skeleton at room temperature. The finite element (FE) analysis has been performed using ANSYS, in order to determine the elongation of the FA skeleton as well as the location of max. stress and stresses developed in axial direction under tensile load of 9800 N or 2 g being the FA handling or lifting load [Y. Zhang et al., Fuel Assembly Design Report, SNERDI, China, 1994]. The FE model of grids, guide thimbles with dash-pots and flow holes has been developed using Shell 181. It has been observed that FA skeleton elongation values obtained through FE analysis and experiment are comparable and show linear behaviors. Moreover, the values of stresses obtained at different locations of the guide thimbles are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Therefore, validation of the FE methodology is confirmed. The values of stresses are less than the design limit of the materials used for the grid and the guide thimble. Therefore, the structural integrity criterion of CHASNUPP-1 FA skeleton is fulfilled safely.

Scholarly Research Program in Fuel Analysis and Combustion Research

DTIC Science & Technology

1993-02-01

Public reporting burden for this collection of information is es•tmated to average I hour per response, ilnduding the time fo," reviwing ...Thermal Oxidative Flask Test 45 9. Advanced Fuel System Configuration Descent Condition 57 10. TGPGC for n-Alkane Mixture 63 11. Hierarchical Cluster ...will include all analytical data, data analysis conclusions, recommendations and rationale. 16 a& k : 05 Titl: Development of Test Cell Assemblies for

Interim status report on lead-cooled fast reactor (LFR) research and development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

2008-03-31

This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigationmore » of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.« less

Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ellis, Ronald James

The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) duringmore » cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.« less

Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

NASA Astrophysics Data System (ADS)

Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

2017-09-01

Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andress, D.; Joy, D.S.; McLeod, N.B.

The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elementsmore » as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.« less

Fuel injection assembly for use in turbine engines and method of assembling same

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2015-03-24

A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes a plurality of tube assemblies, wherein each of the tube assemblies includes an upstream portion and a downstream portion. Each tube assembly includes a plurality of tubes that extend from the upstream portion to the downstream portion or from the upstream portion through the downstream portion. At least one injection system is coupled to at least one tube assembly of the plurality of tube assemblies. The injection system includes a fluid supply member that extends from a fluid source to the downstream portion of the tube assembly. The fluid supply member includes a first end portion located in the downstream portion of the tube assembly, wherein the first end portion has at least one first opening for channeling fluid through the tube assembly to facilitate reducing a temperature therein.

Sputter-deposited fuel cell membranes and electrodes

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R. (Inventor); Jeffries-Nakamura, Barbara (Inventor); Chun, William (Inventor); Ruiz, Ron P. (Inventor); Valdez, Thomas I. (Inventor)

2001-01-01

A method for preparing a membrane for use in a fuel cell membrane electrode assembly includes the steps of providing an electrolyte membrane, and sputter-depositing a catalyst onto the electrolyte membrane. The sputter-deposited catalyst may be applied to multiple sides of the electrolyte membrane. A method for forming an electrode for use in a fuel cell membrane electrode assembly includes the steps of obtaining a catalyst, obtaining a backing, and sputter-depositing the catalyst onto the backing. The membranes and electrodes are useful for assembling fuel cells that include an anode electrode, a cathode electrode, a fuel supply, and an electrolyte membrane, wherein the electrolyte membrane includes a sputter-deposited catalyst, and the sputter-deposited catalyst is effective for sustaining a voltage across a membrane electrode assembly in the fuel cell.

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph; Fugate, Michael Lynn; Trellue, Holly Renee

2016-10-28

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly tomore » confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).« less

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE PAGES

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; ...

2015-12-01

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

Fabrication of capsule assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1973-01-01

Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.

Nuclear core and fuel assemblies

DOEpatents

Downs, Robert E.

1981-01-01

A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, W.E.; Trapp, T.J.

1983-06-10

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, William E.; Trapp, Turner J.

1985-01-01

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly †

PubMed Central

De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

2016-01-01

Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

DOEpatents

Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

2007-10-02

A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

Separator assembly for use in spent nuclear fuel shipping cask

DOEpatents

Bucholz, James A.

1983-01-01

A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

DOE PAGES

Williams, M. L.; Wiarda, D.; Ilas, G.; ...

2014-06-15

Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

Cerium migration during PEM fuel cell assembly and operation

DOE PAGES

Baker, Andrew M.; Torraco, Dennis; Judge, Elizabeth J.; ...

2015-09-14

Cerium migration between PEM fuel cell components is influenced by potential-driven mobility, ionic diffusion, and gradients in water content. These factors were investigated in ex situ experiments and in operating fuel cells. Potential-induced migration was measured ex situ in hydrated window cells. Cerium-containing MEAs were also fabricated and tested under ASTs. MEA disassembly and subsequent XRF analysis were used to observe rapid cerium migration during cell assembly and operation. During MEA hot pressing, humidification, and low RH operation at OCV, ionic diffusion causes uniform migration from the membrane into the catalyst layers. During high RH operation at OCV, in-plane ceriummore » gradients arise due to variations in water content. These gradients may diminish the scavenging efficacy of cerium by reducing its proximity to generated radicals.« less

Current conducting end plate of fuel cell assembly

DOEpatents

Walsh, Michael M.

1999-01-01

A fuel cell assembly has a current conducting end plate with a conductive body formed integrally with isolating material. The conductive body has a first surface, a second surface opposite the first surface, and an electrical connector. The first surface has an exposed portion for conducting current between a working section of the fuel cell assembly and the electrical connector. The isolating material is positioned on at least a portion of the second surface. The conductive body can have support passage(s) extending therethrough for receiving structural member(s) of the fuel cell assembly. Isolating material can electrically isolate the conductive body from the structural member(s). The conductive body can have service passage(s) extending therethrough for servicing one or more fluids for the fuel cell assembly. Isolating material can chemically isolate the one or more fluids from the conductive body. The isolating material can also electrically isolate the conductive body from the one or more fluids.

Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

DOEpatents

Oosterkamp, Willem Jan; Marquino, Wayne

1999-01-05

A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

Burnable absorber arrangement for fuel bundle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowther, R.L.; Townsend, D.B.

1986-12-16

This patent describes a boiling water reactor core whose operation is characterized by a substantial proportion of steam voids with concomitantly reduced moderation toward the top of the core when the reactor is in its hot operating condition. The reduced moderation leads to slower burnup and greater conversion ratio in an upper core region so that when the reactor is in its cold shut down condition the resulting relatively increased moderation in the upper core region is accompanied by a reactivity profile that peaks in the upper core region. A fuel assembly is described comprising; a component of fissile materialmore » distributed over a substantial axial extent of the fuel assembly; and a component of neutron absorbing material having an axial distribution characterized by an enhancement in an axial zone of the fuel assembly, designated the cold shutdown control zone, corresponding to at least a portion of the axial region of the core when the cold shutdown reactivity peaks. The aggregate amount of neutron absorbing material in the cold shutdown zone of the fuel assembly is greater than the aggregate amount of neutron absorbing material in the axial zones of the fuel assembly immediately above and immediately below the cold shutdown control zone whereby the cold shutdown reactivity peak is reduced relative to the cold shutdown reactivity in the zones immediately above and immediately below the cold shutdown control zone. The cold shutdown zone has an axial extent measured from the bottom of the fuel assembly in the range between 68-88 percent of the height of the fissile material in the fuel assembly.« less

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trahan, Alexis Chanel

New nondestructive assay techniques are sought to better characterize spent nuclear fuel. One of the NDA instruments selected for possible deployment is differential die-away self-interrogation (DDSI). The proposed DDSI approach for spent fuel assembly assay utilizes primarily the spontaneous fission and (α, n) neutrons in the assemblies as an internal interrogating radiation source. The neutrons released in spontaneous fission or (α,n) reactions are thermalized in the surrounding water and induce fission in fissile isotopes, thereby creating a measurable signal from isotopes of interest that would be otherwise difficult to measure. The DDSI instrument employs neutron coincidence counting with 3He tubesmore » and list-mode-based data acquisition to allow for production of Rossi-alpha distributions (RADs) in post-processing. The list-mode approach to data collection and subsequent construction of RADs has expanded the analytical possibilities, as will be demonstrated throughout this thesis. One of the primary advantages is that the measured signal in the form of a RAD can be analyzed in its entirety including determination of die-away times in different time domains. This capability led to the development of the early die-away method, a novel leakage multiplication determination method which is tested throughout the thesis on different sources in simulation space and fresh fuel experiments. The early die-away method is a robust, accurate, improved method of determining multiplication without the need for knowledge of the (α,n) source term. The DDSI technique and instrument are presented along with the many novel capabilities enabled by and discovered through RAD analysis. Among the new capabilities presented are the early die-away method, total plutonium content determination, and highly sensitive missing pin detection. Simulation of hundreds of different spent and fresh fuel assemblies were used to develop the analysis algorithms and the techniques were tested on a variety of spontaneous fission-driven fresh fuel assemblies at Los Alamos National Laboratory and the BeRP ball at the Nevada National Security Site. The development of the new, improved analysis and characterization methods with the DDSI instrument makes it a viable technique for implementation in a facility to meet material control and safeguards needs.« less

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

2016-05-01

This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

2016-05-12

This report describes the third set of tests (the “DCL a shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

DOEpatents

Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ

2009-12-22

The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

Fuel cell design and assembly

DOEpatents

Myerhoff, Alfred

1984-01-01

The present invention is directed to a novel bipolar cooling plate, fuel cell design and method of assembly of fuel cells. The bipolar cooling plate used in the fuel cell design and method of assembly has discrete opposite edge and means carried by the plate defining a plurality of channels extending along the surface of the plate toward the opposite edges. At least one edge of the channels terminates short of the edge of the plate defining a recess for receiving a fastener.

Advanced techniques for repair of irradiated PWR fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knaab, H.; Westphal, M.

Kraftwerk union has recently designed and built a portable repair unit for use in nuclear power plants for repair of defective fuel assemblies where space limitations do not allow permanent installation of repair equipment. This new equipment is designed to be easily disassembled and decontaminated. The main component of the equipment is the fuel assembly reconstitution unit (FARU) which is placed on the floor of the spent fuel pool. The use of the FARU is described in the paper.

Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Da Cruz, D. F.; Rochman, D.; Koning, A. J.

2012-07-01

This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {supmore » 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)« less

Fuel nozzle assembly

DOEpatents

Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

2011-08-30

A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

Internal reforming fuel cell assembly with simplified fuel feed

DOEpatents

Farooque, Mohammad; Novacco, Lawrence J.; Allen, Jeffrey P.

2001-01-01

A fuel cell assembly in which fuel cells adapted to internally reform fuel and fuel reformers for reforming fuel are arranged in a fuel cell stack. The fuel inlet ports of the fuel cells and the fuel inlet ports and reformed fuel outlet ports of the fuel reformers are arranged on one face of the fuel cell stack. A manifold sealing encloses this face of the stack and a reformer fuel delivery system is arranged entirely within the region between the manifold and the one face of the stack. The fuel reformer has a foil wrapping and a cover member forming with the foil wrapping an enclosed structure.

Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, W.J.; Husser, D.L.; Mohr, T.C.

2004-02-04

New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developedmore » to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.« less

Combustor assembly for use in a turbine engine and methods of assembling same

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2013-05-14

A fuel nozzle assembly for use with a turbine engine is described herein. The fuel nozzle assembly includes a plurality of fuel nozzles positioned within an air plenum defined by a casing. Each of the plurality of fuel nozzles is coupled to a combustion liner defining a combustion chamber. Each of the plurality of fuel nozzles includes a housing that includes an inner surface that defines a cooling fluid plenum and a fuel plenum therein, and a plurality of mixing tubes extending through the housing. Each of the mixing tubes includes an inner surface defining a flow channel extending between the air plenum and the combustion chamber. At least one mixing tube of the plurality of mixing tubes including at least one cooling fluid aperture for channeling a flow of cooling fluid from the cooling fluid plenum to the flow channel.

Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

DOEpatents

Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

1985-01-01

A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

DOEpatents

Oosterkamp, W.J.; Marquino, W.

1999-01-05

A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

Hybrid Gama Emission Tomography (HGET): FY16 Annual Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Erin A.; Smith, Leon E.; Wittman, Richard S.

2017-02-01

Current International Atomic Energy Agency (IAEA) methodologies for the verification of fresh low-enriched uranium (LEU) and mixed oxide (MOX) fuel assemblies are volume-averaging methods that lack sensitivity to individual pins. Further, as fresh fuel assemblies become more and more complex (e.g., heavy gadolinium loading, high degrees of axial and radial variation in fissile concentration), the accuracy of current IAEA instruments degrades and measurement time increases. Particularly in light of the fact that no special tooling is required to remove individual pins from modern fuel assemblies, the IAEA needs new capabilities for the verification of unirradiated (i.e., fresh LEU and MOX)more » assemblies to ensure that fissile material has not been diverted. Passive gamma emission tomography has demonstrated potential to provide pin-level verification of spent fuel, but gamma-ray emission rates from unirradiated fuel emissions are significantly lower, precluding purely passive tomography methods. The work presented here introduces the concept of Hybrid Gamma Emission Tomography (HGET) for verification of unirradiated fuels, in which a neutron source is used to actively interrogate the fuel assembly and the resulting gamma-ray emissions are imaged using tomographic methods to provide pin-level verification of fissile material concentration.« less

Fuel-Mediated Transient Clustering of Colloidal Building Blocks.

PubMed

van Ravensteijn, Bas G P; Hendriksen, Wouter E; Eelkema, Rienk; van Esch, Jan H; Kegel, Willem K

2017-07-26

Fuel-driven assembly operates under the continuous influx of energy and results in superstructures that exist out of equilibrium. Such dissipative processes provide a route toward structures and transient behavior unreachable by conventional equilibrium self-assembly. Although perfected in biological systems like microtubules, this class of assembly is only sparsely used in synthetic or colloidal analogues. Here, we present a novel colloidal system that shows transient clustering driven by a chemical fuel. Addition of fuel causes an increase in hydrophobicity of the building blocks by actively removing surface charges, thereby driving their aggregation. Depletion of fuel causes reappearance of the charged moieties and leads to disassembly of the formed clusters. This reassures that the system returns to its initial, equilibrium state. By taking advantage of the cyclic nature of our system, we show that clustering can be induced several times by simple injection of new fuel. The fuel-mediated assembly of colloidal building blocks presented here opens new avenues to the complex landscape of nonequilibrium colloidal structures, guided by biological design principles.

SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

DOEpatents

Schluderberg, D.C.; Ryon, J.W.

1962-05-01

A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

Air breathing direct methanol fuel cell

DOEpatents

Ren, Xiaoming; Gottesfeld, Shimshon

2002-01-01

An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source. Water loss from the cell is minimized by making the conductive cathode assembly hydrophobic and the conductive anode assembly hydrophilic.

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

Upgraded HFIR Fuel Element Welding System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sease, John D

2010-02-01

The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. Inmore » recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.« less

Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, Fariz Abdul; Lee, John C.; Franceschini, Fausto

2012-07-01

As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning themore » legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and burn TRU in a thermal spectrum, while satisfying top-level operational and safety constraints. Various assembly designs have been proposed to assess the TRU burning potential of Th-based fuel in PWRs. In addition to typical homogeneous loading patterns, heterogeneous configurations exploiting the breeding potential of thorium to enable multiple cycles of TRU irradiation and burning have been devised. The homogeneous assembly design, with all pins featuring TRU in Th, has the benefit of a simple loading pattern and the highest rate of TRU transmutation, but it can be used only for a few cycles due to the rapid rise in the TRU content of the recycled fuel, which challenges reactivity control, safety coefficients and fuel handling. Due to its simple loading pattern, such assembly design can be used as the first step of Th implementation, achieving up to 3 times larger TRU transmutation rate than conventional U-MOX, assuming same fraction of MOX assemblies in the core. As the next step in thorium implementation, heterogeneous assemblies featuring a mixed array of Th-U and Th-U-TRU pins, where the U is in-bred from Th, have been proposed. These designs have the potential to enable burning an external supply of TRU through multiple cycles of irradiation, recovery (via reprocessing) and recycling of the residual actinides at the end of each irradiation cycle. This is achieved thanks to a larger breeding of U from Th in the heterogeneous assemblies, which reduces the TRU supply and thus mitigates the increase in the TRU core inventory for the multi-recycled fuel. While on an individual cycle basis the amount of TRU burned in the heterogeneous assembly is reduced with respect to the homogeneous design, TRU burning rates higher than single-pass U-MOX fuel can still be achieved, with the additional benefits of a multi-cycle transmutation campaign recycling all TRU isotopes. Nitride fuel, due its higher density and U breeding potential, together with its better thermal properties, ideally suits the objectives and constraints of the heterogeneous assemblies. However, significant technological advancements must be made before nitride fuels can be employed in an LWR: its water resistance needs to be improved and a viable technology to enrich N in N-15 must be devised. Moreover, for the nitride heterogeneous configurations examined in this study, the enhancement in TRU burning performance is achieved not only by replacing oxide with nitride fuel, but also by increasing the fuel rod size. This latter modification, allowed by the high thermal conductivity of nitride fuel, leads however to a very tight lattice, which may challenge reactor coolant pumps and assembly hold-down mechanisms, the former through an increase in core pressure drop and the latter through an increase in assembly lift-off forces. To alleviate these issues, while still achieving the large fuel-to-moderator ratios resulting from using tight lattices, wire wraps could be used in place of grid spacers. For tight lattices, typical grid spacers are hard to manufacture and their replacement with wire wraps is known to allow for a pressure drop reduction by at least 2 times. The studies, while certainly very preliminary, provide a starting point to devise an optimum strategy for TRU transmutation in Th-based PWR fuel. The viability of the scheme proposed depends on the timely phasing in of the associated technologies, with proper lead time and to solve the many challenges. These challenges are certainly substantial, and make the current once-through U-based scheme pursued in the US by far a more practical (and cheaper) option. However, when compared to other transmutation schemes, the proposed one has arguably similar challenges and unknowns with potentially bigger rewards. (authors)« less

ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wieselquist, William A.; Thompson, Adam B.; Bowman, Stephen M.

2016-04-01

Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process datamore » to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.« less

75 FR 901 - Airworthiness Directives; Honeywell International Inc. ALF502 Series and LF507 Series Turbofan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-07

.... ALF502 series and LF507 series turbofan engines with certain fuel manifold assemblies installed. That AD... of certain part number (P/N) fuel manifold assemblies for cracks, and replacement of cracked fuel manifolds with serviceable manifolds. This AD continues to require inspecting those fuel manifolds for...

77 FR 72200 - Airworthiness Directives; The Boeing Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-05

... correctly on the engine fuel feed manifold couplings. This AD also requires inspecting the assembly of the engine fuel feed manifold rigid and full flexible couplings. This AD was prompted by reports of fuel leaks due to improperly assembled engine fuel feed manifold couplings. We are issuing this AD to detect...

Fuel cell cooler-humidifier plate

DOEpatents

Vitale, Nicholas G.; Jones, Daniel O.

2000-01-01

A cooler-humidifier plate for use in a proton exchange membrane (PEM) fuel cell stack assembly is provided. The cooler-humidifier plate combines functions of cooling and humidification within the fuel cell stack assembly, thereby providing a more compact structure, simpler manifolding, and reduced reject heat from the fuel cell. Coolant on the cooler side of the plate removes heat generated within the fuel cell assembly. Heat is also removed by the humidifier side of the plate for use in evaporating the humidification water. On the humidifier side of the plate, evaporating water humidifies reactant gas flowing over a moistened wick. After exiting the humidifier side of the plate, humidified reactant gas provides needed moisture to the proton exchange membranes used in the fuel cell stack assembly. The invention also provides a fuel cell plate that maximizes structural support within the fuel cell by ensuring that the ribs that form the boundaries of channels on one side of the plate have ends at locations that substantially correspond to the locations of ribs on the opposite side of the plate.

Automatically closing swing gate closure assembly

DOEpatents

Chang, Shih-Chih; Schuck, William J.; Gilmore, Richard F.

1988-01-01

A swing gate closure assembly for nuclear reactor tipoff assembly wherein the swing gate is cammed open by a fuel element or spacer but is reliably closed at a desired closing rate primarily by hydraulic forces in the absence of a fuel charge.

Disposal of spent fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blomeke, J O; Ferguson, D E; Croff, A G

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed betweenmore » the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom.« less

End plate assembly having a two-phase fluid-filled bladder and method for compressing a fuel cell stack

DOEpatents

Carlstrom, Jr., Charles M.

2001-01-01

An end plate assembly is disclosed for use in a fuel cell assembly in which the end plate assembly includes a housing having a cavity, and a bladder receivable in the cavity and engageable with the fuel cell stack. The bladder includes a two-phase fluid having a liquid portion and a vapor portion. Desirably, the two-phase fluid has a vapor pressure between about 100 psi and about 600 psi at a temperature between about 70 degrees C. to about 110 degrees C.

Used fuel rail shock and vibration testing options analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ross, Steven B.; Best, Ralph E.; Klymyshyn, Nicholas A.

2014-09-25

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

Miniature ceramic fuel cell

DOEpatents

Lessing, Paul A.; Zuppero, Anthony C.

1997-06-24

A miniature power source assembly capable of providing portable electricity is provided. A preferred embodiment of the power source assembly employing a fuel tank, fuel pump and control, air pump, heat management system, power chamber, power conditioning and power storage. The power chamber utilizes a ceramic fuel cell to produce the electricity. Incoming hydro carbon fuel is automatically reformed within the power chamber. Electrochemical combustion of hydrogen then produces electricity.

78 FR 76328 - Application for Renewal of Special Nuclear Material License SNM-2014 From Tennessee Valley...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-17

... SNM in the form of fully-assembled fuel assemblies that would later form the initial reactor core of WBN2. The SNM in the fuel assemblies is enriched up to 5% in the isotope U-235. The fresh fuel... received the initial core for WBN2. The NRC has not yet issued the OL for the Unit 2 reactor. The...

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

NASA Astrophysics Data System (ADS)

Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.; Menlove, Howard O.; Flaska, Marek; Pozzi, Sara A.

2017-07-01

The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrument that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.

Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.

The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. Thus the NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrumentmore » that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.« less

Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

DOE PAGES

Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.; ...

2017-01-05

The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. Thus the NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrumentmore » that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.« less

METHOD AND MEANS FOR SUPPORTING REACTOR FUEL CONTAINERS IN AN ASSEMBLY

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.; Coombs, C.A.

1962-12-11

This patent relates to means for supporting fuelcontaining tubes in an assembly which include grid means at either end of the fuel element assembly antl improved grid means intermediate of the ends to provide support against lateral displacement. (AEC)

TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.

Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment whilemore » safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional TREAT fuel elements to facilitate the experiment will not inhibit the ability to successfully simulate a RIA for the 2-pin or 3-pin bundle. This new water loop design leaves room for accommodating a larger fuel pin bundle than previously analyzed. The 7-pin fuel bundle in a hexagonal array with similar spacing of fuel pins in a SFR fuel assembly was considered the minimum needed for one central fuel pin to encounter the most correct thermal conditions. The 9-rod fuel bundle in a square array similar in spacing to pins in a LWR fuel assembly would be considered the LWR equivalent. MCNP analysis conducted on a preliminary LWR 9-rod bundle design shows that sufficient energy deposition into the central pin can be achieved well within range to investigate fuel and cladding performance in a simulated RIA. This is achieved by surrounding the flow channel with an additional annulus of water. Findings also show that a highly significant increase in TREAT to specimen power coupling factor (PCF) within the central pin can be achieved by surrounding the experiment with one to two rings of TREAT upgrade fuel assemblies. The experiment design holds promise for the performance evaluation of PWR fuel at extremely high burnup under similar reactor environment conditions.« less

Solid oxide fuel cell generator with removable modular fuel cell stack configurations

DOEpatents

Gillett, J.E.; Dederer, J.T.; Zafred, P.R.; Collie, J.C.

1998-04-21

A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack. 8 figs.

Solid oxide fuel cell generator with removable modular fuel cell stack configurations

DOEpatents

Gillett, James E.; Dederer, Jeffrey T.; Zafred, Paolo R.; Collie, Jeffrey C.

1998-01-01

A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack.

Calculated Drag of an Aerial Refueling Assembly Through Airplane Performance Analysis

NASA Technical Reports Server (NTRS)

Vachon, Jake; Ray, Ronald; Calianno, Carl

2004-01-01

This viewgraph document reviews NASA Dryden's work on Aerial refueling, with specific interest in calculating the drag of the refueling system. The aerodynamic drag of an aerial refueling assembly was calculated during the Automated Aerial Refueling project at the NASA Dryden Flight Research Center. An F/A-18A airplane was specially instrumented to obtain accurate fuel flow measurements and to determine engine thrust

96. SEED 1 FUEL ASSEMBLY FROM LOCATION L9 BEING REMOVED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

96. SEED 1 FUEL ASSEMBLY FROM LOCATION L-9 BEING REMOVED FROM REACTOR VESSEL BY MEANS OF FUEL EXTRACTION CRANE, JANUARY 7, 1960 - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) SHOWING CRANE ASSEMBLY ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) SHOWING CRANE ASSEMBLY FOR TRANSFER PIT. INL PHOTO NUMBER NRTS-51-2404. Unknown Photographer, 5/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Nuclear reactor composite fuel assembly

DOEpatents

Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

1980-01-01

A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

Passive Safety Features Evaluation of KIPT Neutron Source Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhong, Zhaopeng; Gohar, Yousry

2016-06-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. It will be used to conduct basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The facility has an electron accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assemblymore » is composed of WWR-M2 - Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The subcritical assembly is seated in a water tank, which is a part of the primary cooling loop. During normal operation, the water coolant operates at room temperature and the total facility power is ~300 KW. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B-VII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight water temperature increase and the corresponding water density decrease produce large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. The reactivity variation due to an empty fuel position filled by water during the fuel loading process is examined. Also, the loading mistakes of removing beryllium reflector assemblies and replacing them with dummy assemblies were analyzed. In all these circumstances, the reactivity change results do not cause any safety concerns.« less

Air breathing direct methanol fuel cell

DOEpatents

Ren, Xiaoming

2002-01-01

An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source.

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE PAGES

Hu, Jianwei; Gauld, Ian C.

2014-12-01

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Gauld, Ian C.

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-11-07

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-01-01

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bi, G.; Liu, C.; Si, S.

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis ofmore » reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)« less

A custom-tailored FAMOS burn-up meter for VVER 440 fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simon, G.G.; Golochtchapov, S.; Glazov, A.G.

1995-12-31

The FAMOS fuel assembly monitoring system had been originally developed for monitoring irradiated fuel assemblies of the Karlsruhe Nuclear Research Center concentrating on neutron detection systems for special applications.The measurements in the past had demonstrated that FAMOS can perform precise measurements to control or measure with accuracy the main physical parameters of spent fuel. The FAMOS 3 system is specialized for burn-up determination of fuel assemblies. Thus it is possible to take into account the burn-up for the purposes of storage and transportation. The Kola NPP VVER 440 requirements necessitated developing an especially adopted FAMOS 3 system. In addition tomore » the passive neutron measurement, a gross gamma detection and a boron concentration monitoring system are implemented. The new system was constructed as well as tested in laboratory experiments. The monitoring system has been delivered to the customer and is ready for use.« less

Independent Orbiter Assessment (IOA): Analysis of the electrical power generation/fuel cell powerplant subsystem

NASA Technical Reports Server (NTRS)

Brown, K. L.; Bertsch, P. J.

1986-01-01

Results of the Independent Orbiter Assessment (IOA) of the Failure Modes and Effects Analysis (FMEA) and Critical Items List (CIL) are presented. The IOA approach features a top-down analysis of the hardware to determine failure modes, criticality, and potential critical items. To preserve independence, this analysis was accomplished without reliance upon the results contained within the NASA FMEA/CIL documentation. This report documents the independent analysis results corresponding to the Orbiter Electrical Power Generation (EPG)/Fuel Cell Powerplant (FCP) hardware. The EPG/FCP hardware is required for performing functions of electrical power generation and product water distribution in the Orbiter. Specifically, the EPG/FCP hardware consists of the following divisions: (1) Power Section Assembly (PSA); (2) Reactant Control Subsystem (RCS); (3) Thermal Control Subsystem (TCS); and (4) Water Removal Subsystem (WRS). The IOA analysis process utilized available EPG/FCP hardware drawings and schematics for defining hardware assemblies, components, and hardware items. Each level of hardware was evaluated and analyzed for possible failure modes and effects. Criticality was assigned based upon the severity of the effect for each failure mode.

Transfer of fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vuckovich, M.; Burkett, J. P.; Sallustio, J.

1984-12-11

Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If themore » strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint. This event occurs when a fuel assembly being deposited contacts the bottom of its slot or an obstruction in, or at the entry to the slot. To preclude lateral movement and possible damage to a fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge exceets the lov-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than the slack-line set-point. This occurs when the lines has been set in slack-line setting. When the line is tensioned after slack-li ne setting, the traverse drive remains enabled only if the line has been disconnected from the fuel assembly.« less

Evaluation of Cask Drop Criticality Issues at K Basin

DOE Office of Scientific and Technical Information (OSTI.GOV)

GOLDMANN, L.H.

An analysis of ability of Multi-canister Overpack (MCO) to withstand drops at K Basin without exceeding the criticality design requirements. Report concludes the MCO will function acceptably. The spent fuel currently residing in the 105 KE and 105 KW storage basins will be placed in fuel storage baskets which will be loaded into the MCO cask assembly. During the basket loading operations the MCO cask assembly will be positioned near the bottom of the south load out pit (SLOP). The loaded MCO cask will be lifted from the SLOP transferred to the transport trailer and delivered to the Cold Vacuummore » Drying Facility (CVDF). In the wet condition there is a potential for criticality problems if significant changes in the designed fuel configurations occur. The purpose of this report is to address structural issues associated with criticality design features for MCO cask drop accidents in the 105 KE and 105 KW facilities.« less

Polymer electrolyte membrane assembly for fuel cells

NASA Technical Reports Server (NTRS)

Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

2002-01-01

An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

Polymer electrolyte membrane assembly for fuel cells

NASA Technical Reports Server (NTRS)

Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

2000-01-01

An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

78 FR 9796 - Airworthiness Directives; Cessna Aircraft Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-12

... against the right steering tube assembly during rudder pedal actuation. This AD requires you to install... between the fuel return line assembly and the steering tube assembly and clearance between the fuel return...://www.cessnasupport.com . You may review copies of the referenced service information at the FAA, Small...

77 FR 50054 - Airworthiness Directives; Cessna Aircraft Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-20

... rubbing against the right steering tube assembly during full rudder pedal actuation. This proposed AD would require you to inspect the fuel return line assembly for chafing; replace the fuel return line... right steering tube assembly and the airplane structure; and adjustment as necessary. We are proposing...

77 FR 72250 - Airworthiness Directives; Cessna Aircraft Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-05

... rubbing against the right steering tube assembly during rudder pedal actuation. This proposed AD would require you to install the forward and aft fuel return line support clamps and brackets; inspect for a minimum clearance between the fuel return line assembly and the steering tube assembly and clearance...

CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

DOEpatents

Weems, S.J.

1963-09-24

>A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1991-05-21

This patent describes an improved fuel assembly for a boiling water reactor. It comprises: a fuel channel; a lower tie plate; an upper tie plate; the lower tie plate and the upper tie plate defining a two-dimensional matrix; at least one water rod the fuel rods being partial length rods.

75 FR 3876 - Mark Edward Leyse; Receipt of Petition for Rulemaking

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-25

... (assembly) severe fuel damage experiments. The petitioner also requests that the NRC promulgate a regulation... aware that data from multi-rod (assembly) severe fuel damage experiments indicates that the current... fuel damage experiments indicates that the current peak cladding temperature limit contained in 10 CFR...

FY 2016 Status Report: Documentation of All CIRFT Data including Hydride Reorientation Tests (Draft M2)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Jiang, Hao

The first portion of this report provides a detailed description of fiscal year (FY) 2015 test result corrections and analysis updates based on FY 2016 updates to the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) program methodology, which is used to evaluate the vibration integrity of spent nuclear fuel (SNF) under normal conditions of transport (NCT). The CIRFT consists of a U-frame test setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages connecting to a universal testing machine. The curvature SNF rod bending is obtained through a three-point deflection measurementmore » method. Three linear variable differential transformers (LVDTs) are clamped to the side connecting plates of the U-frame and used to capture deformation of the rod. The second portion of this report provides the latest CIRFT data, including data for the hydride reorientation test. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNF samples into a single zone. A detailed examination revealed that, at the same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), LMK, and mixed uranium-plutonium oxide (MOX). Just looking at the strain, LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The third portion of this report provides finite element analysis (FEA) dynamic deformation simulation of SNF assemblies . In a horizontal layout under NCT, the fuel assembly’s skeleton, which is formed by guide tubes and spacer grids, is the primary load bearing apparatus carrying and transferring vibration loads within an SNF assembly. These vibration loads include interaction forces between the SNF assembly and the canister basket walls. Therefore, the integrity of the guide tubes and spacer grids critically affects the vibration intensity of the fuel assembly during transport and must be considered when developing the multipurpose purpose canister (MPC) design for safe SNF transport.« less

General view of a Space Shuttle Main Engine (SSME) mounted ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

General view of a Space Shuttle Main Engine (SSME) mounted on an SSME engine handler, taken in the SSME Processing Facility at Kennedy Space Center. The most prominent features of the engine assembly in this view are the Low-Pressure Fuel Turbopump Discharge Duct looping around the right side and underneath the assembly, the High-Pressure Fuel Turbopump located on the lower left portion of the assembly, the Engine Controller and Main Fuel Valve Hydraulic Actuator located on the upper portion of the assembly and the Low-Pressure Oxidizer Turbopump Discharge Duct at the top of the engine assembly in this view. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

PWR integral tie plate and locking mechanism

DOE Office of Scientific and Technical Information (OSTI.GOV)

Flora, B.S.; Osborne, J.L.

1980-08-26

A locking mechanism for securing an upper tie plate to the tie rods of a nuclear fuel bundle is described. The mechanism includes an upper tie plate assembly and locking sleeves fixed to the ends of the tie rods. The tie plate is part of the upper tie plate assembly and is secured to the fuel bundle by securing the entire upper tie plate assembly to the locking sleeves fixed to the tie rods. The assembly includes, in addition to the tie plate, locking nuts for engaging the locking sleeves, retaining sleeves to operably connect the locking nuts to themore » assembly, a spring biased reaction plate to restrain the locking nuts in the locked position and a means to facilitate the removal of the entire assembly as a unit from the fuel bundle.« less

Portable instrument for inspecting irradiated nuclear fuel assemblies

DOEpatents

Nicholson, Nicholas; Dowdy, Edward J.; Holt, David M.; Stump, Jr., Charles J.

1985-01-01

A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

Method and apparatus for assembling solid oxide fuel cells

DOEpatents

Szreders, B.E.; Campanella, N.

1988-05-11

This invention relates generally to solid oxide fuel power generators and is particularly directed to improvements in the assembly and coupling of solid oxide fuel cell modules. A plurality of jet air tubes are supported and maintained in a spaced matrix array by a positioning/insertion assembly for insertion in respective tubes of a solid oxide fuel cell (SOFC) in the assembly of an SOFC module. The positioning/insertion assembly includes a plurality of generally planar, elongated, linear vanes which are pivotally mounted at each end thereof to a support frame. A rectangular compression assembly of adjustable size is adapted to receive and squeeze a matrix of SOFC tubes so as to compress the inter-tube nickel felt conductive pads which provide series/parallel electrical connection between adjacent SOFCs, with a series of increasingly larger retainer frames used to maintain larger matrices of SOFC tubes in position. Expansion of the SOFC module housing at the high operating temperatures of the SOFC is accommodated by conductive, flexible, resilient expansion, connector bars which provide support and electrical coupling at the top and bottom of the SOFC module housing. 17 figs.

Morphological features (defects) in fuel cell membrane electrode assemblies

NASA Astrophysics Data System (ADS)

Kundu, S.; Fowler, M. W.; Simon, L. C.; Grot, S.

Reliability and durability issues in fuel cells are becoming more important as the technology and the industry matures. Although research in this area has increased, systematic failure analysis, such as a failure modes and effects analysis (FMEA), are very limited in the literature. This paper presents a categorization scheme of causes, modes, and effects related to fuel cell degradation and failure, with particular focus on the role of component quality, that can be used in FMEAs for polymer electrolyte membrane (PEM) fuel cells. The work also identifies component defects imparted on catalyst-coated membranes (CCM) by manufacturing and proposes mechanisms by which they can influence overall degradation and reliability. Six major defects have been identified on fresh CCM materials, i.e., cracks, orientation, delamination, electrolyte clusters, platinum clusters, and thickness variations.

PLATES WITH OXIDE INSERTS

DOEpatents

West, J.M.; Schumar, J.F.

1958-06-10

Planar-type fuel assemblies for nuclear reactors are described, particularly those comprising fuel in the oxide form such as thoria and urania. The fuel assembly consists of a plurality of parallel spaced fuel plate mennbers having their longitudinal side edges attached to two parallel supporting side plates, thereby providing coolant flow channels between the opposite faces of adjacent fuel plates. The fuel plates are comprised of a plurality of longitudinally extending tubular sections connected by web portions, the tubular sections being filled with a plurality of pellets of the fuel material and the pellets being thermally bonded to the inside of the tubular section by lead.

Hatch assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, J.R.; Hardin, R.T. Jr.

1987-07-07

This patent describes a nuclear reactor installation including means defining a fuel handling area and means defining a containment area separated from the fuel handling area and including a refuelling cavity; the improvement comprising: (a) a fuel transfer tube connecting the refuelling cavity with the fuel handling area; the fuel transfer tube having a first end in the fuel handling area and a second end in the refueling cavity; (b) valve means for opening and closing the first end; and (c) a hatch assembly mounted on the second end; the hatch assembly including (1) a hatch ring affixed to themore » fuel transfer tube at the second end the hatch ring has an integral annular seat surrounded by the hatch ring and defines a hatch opening in the second end of the fuel transfer tube; (2) a hatch cover adapts to be positioned on the annular seat for covering the hatch opening; (3) latching units are supported on the hatch ring about the hatch opening, each latching unit.« less

Recovering tubewise power from three-dimensional nodal kinetics calculation during material relocation in an HWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalimullah; Morris, E.E.; Yang, W.S.

1994-12-31

To analyze severe accidents in some special-purpose heavy-water reactors made of assemblies consisting of a number of coaxial tubes of aluminum-clad U-Al fuel and aluminum-clad neutron-capturing material, a mechanistic model, MARTINS, for tube beatup, melting, and molten material relocation has been developed and integrated with the DIF3D nodal hexagonal-z reactor kinetics and other phenomenological modules. The DIF3D kinetics homogenizes all materials located and computes the total power produced in an axial segment of a fuel assembly. This paper presents an approximate method, used in MARTINS, to calculate the distribution of this total nodal power into the intact fuel and capturingmore » material tubes and the meat-cladding mixtures relocating during tube disruption. The method accounts for the change in intraassembly radial power profile due to assembly geometry change with the progress of segment-by-segment disruption of different tubes. Earlier methods to recover pinwise power from nodal calculation for liquid-metal-cooled reactors and light water reactors (X-Y and hexagonal unit cells) are not practical for a disrupting assembly having material relocation. Figure 1 shows the assembly`s end view, divided into rings for modeling and analysis. A ring is a coolant subchannel plus the outer surrounding tube. The present method for distributing the nodal power consists of two parts: (a) calculation of the relative values of ring-by-ring power per unit uranium mass and power per unit mass of neutron-capturing material in a given assembly segment, and (b) normalization of these relative values such that the total power of all rings (intact tubes and U-Al-Cp meat-cladding mixtures, where Cp implies the neutron-capturing material) equals the DIF3D-calculated nodal power for the assembly axial segment.« less

NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

DOEpatents

McCorkle, W.H.

1960-02-23

BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

High efficiency gas burner

DOEpatents

Schuetz, Mark A.

1983-01-01

A burner assembly provides for 100% premixing of fuel and air by drawing the air into at least one high velocity stream of fuel without power assist. Specifically, the nozzle assembly for injecting the fuel into a throat comprises a plurality of nozzles in a generally circular array. Preferably, swirl is imparted to the air/fuel mixture by angling the nozzles. The diffuser comprises a conical primary diffuser followed by a cusp diffuser.

Detecting pin diversion from pressurized water reactors spent fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Young S.; Sitaraman, Shivakumar

Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and takingmore » the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.« less

Advanced Fuels for Chemical Lasers

DTIC Science & Technology

1975-05-01

age that the analysis is not limited to a single distance from the injector. The four curves in Fig. 86 demonstrate the types of reactions that are...annular wall for this concentric-element-type injector. The rear manifold assembly is made of machined nickel 1late and contains the four - bolt-hole...fabrication k] and assembly of the fully cooled laser combustor components, a series of four tests was cordicted to make preliminary assessment of

Combustor with two stage primary fuel tube with concentric members and flow regulating

DOEpatents

Parker, David Marchant; Whidden, Graydon Lane; Zolyomi, Wendel

1999-01-01

A combustor for a gas turbine having a centrally located fuel nozzle and inner, middle and outer concentric cylindrical liners, the inner liner enclosing a primary combustion zone. The combustor has an air inlet that forms two passages for pre-mixing primary fuel and air to be supplied to the primary combustion zone. Each of the pre-mixing passages has a circumferential array of swirl vanes. A plurality of primary fuel tube assemblies extend through both pre-mixing passages, with each primary fuel tube assembly located between a pair of swirl vanes. Each primary fuel tube assembly is comprised of two tubular members. The first member supplies fuel to the first pre-mixing passage, while the second member, which extends through the first member, supplies fuel to the second pre-mixing passage. An annular fuel manifold is divided into first and second chambers by a circumferentially extending baffle. The proximal end of the first member is attached to the manifold itself while the proximal end of the second member is attached to the baffle. The distal end of the first member is attached directly to the second member at around its mid-point. The inlets of the first and second members are in flow communication with the first and second manifold chambers, respectively. Control valves separately regulate the flow of fuel to the two chambers and, therefore, to the two members of the fuel tube assemblies, thereby allowing the flow of fuel to the first and second pre-mixing passages to be separately controlled.

Analyzing Structural Changes of Fe-N-C Cathode Catalysts in PEM Fuel Cell by Mößbauer Spectroscopy of Complete Membrane Electrode Assemblies.

PubMed

Kramm, Ulrike I; Lefèvre, Michel; Bogdanoff, Peter; Schmeißer, Dieter; Dodelet, Jean-Pol

2014-11-06

The applicability of analyzing by Mößbauer spectroscopy the structural changes of Fe-N-C catalysts that have been tested at the cathode of membrane electrode assemblies in proton exchange membrane (PEM) fuel cells is demonstrated. The Mößbauer characterization of powders of the same catalysts was recently described in our previous publication. A possible change of the iron species upon testing in fuel cell was investigated here by Mößbauer spectroscopy, energy-dispersive X-ray cross-sectional imaging, and neutron activation analysis. Our results show that the absorption probability of γ rays by the iron nuclei in Fe-N-C is strongly affected by the presence of Nafion and water content. A detailed investigation of the effect of an oxidizing treatment (1.2 V) of the non-noble cathode in PEM fuel cell indicates that the observed activity decay is mainly attributable to carbon oxidation causing a leaching of active iron sites hosted in the carbon matrix.

BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

DOE Office of Scientific and Technical Information (OSTI.GOV)

T.L. Lotz

1997-02-15

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercialmore » spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.« less

Sobol's sensitivity analysis for a fuel cell stack assembly model with the aid of structure-selection techniques

NASA Astrophysics Data System (ADS)

Zhang, Wei; Cho, Chongdu; Piao, Changhao; Choi, Hojoon

2016-01-01

This paper presents a novel method for identifying the main parameters affecting the stress distribution of the components used in assembly modeling of proton exchange membrane fuel cell (PEMFC) stack. This method is a combination of an approximation model and Sobol's method, which allows a fast global sensitivity analysis for a set of uncertain parameters using only a limited number of calculations. Seven major parameters, i.e., Young's modulus of the end plate and the membrane electrode assembly (MEA), the contact stiffness between the MEA and bipolar plate (BPP), the X and Y positions of the bolts, the pressure of each bolt, and the thickness of the end plate, are investigated regarding their effect on four metrics, i.e., the maximum stresses of the MEA, BPP, and end plate, and the stress distribution percentage of the MEA. The analysis reveals the individual effects of each parameter and its interactions with the other parameters. The results show that the X position of a bolt has a major influence on the maximum stresses of the BPP and end plate, whereas the thickness of the end plate has the strongest effect on both the maximum stress and the stress distribution percentage of the MEA.

Requirements to the procedure and stages of innovative fuel development

NASA Astrophysics Data System (ADS)

Troyanov, V.; Zabudko, L.; Grachyov, A.; Zhdanova, O.

2016-04-01

According to the accepted current understanding under the nuclear fuel we will consider the assembled active zone unit (Fuel assembly) with its structural elements, fuel rods, pellet column, structural materials of fuel rods and fuel assemblies. The licensing process includes justification of safe application of the proposed modifications, including design-basis and experimental justification of the modified items under normal operating conditions and in violation of normal conditions, including accidents as well. Besides the justification of modified units itself, it is required to show the influence of modifications on the performance and safety of the other Reactor Unit’ and Nuclear Plant’ elements (e.g. burst can detection system, transportation and processing operations during fuel handling), as well as to justify the new standards of fuel storage etc. Finally, the modified fuel should comply with the applicable regulations, which often becomes a very difficult task, if only because those regulations, such as the NP-082-07, are not covered modification issues. Making amendments into regulations can be considered as the only solution, but the process is complicated and requires deep grounds for amendments. Some aspects of licensing new nuclear fuel are considered the example of mixed nitride uranium -plutonium fuel application for the BREST reactor unit.

Storage, transportation and disposal system for used nuclear fuel assemblies

DOEpatents

Scaglione, John M.; Wagner, John C.

2017-01-10

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

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE PAGES

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.; ...

2017-04-28

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Portable instrument for inspecting irradiated nuclear-fuel assemblies in a water-filled storage pond by measurement of induced Cerenkov radiation

DOEpatents

Nicholson, N.; Dowdy, E.J.; Holt, D.M.; Stump, C.J. Jr.

1982-05-13

A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

Fuel cell with electrolyte matrix assembly

DOEpatents

Kaufman, Arthur; Pudick, Sheldon; Wang, Chiu L.

1988-01-01

This invention is directed to a fuel cell employing a substantially immobilized electrolyte imbedded therein and having a laminated matrix assembly disposed between the electrodes of the cell for holding and distributing the electrolyte. The matrix assembly comprises a non-conducting fibrous material such as silicon carbide whiskers having a relatively large void-fraction and a layer of material having a relatively small void-fraction.

Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, J.C.

2001-09-28

The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staffmore » has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs, they do demonstrate that the effect of BPRs is generally well behaved and that independent codes and cross-section libraries predict similar results. The report concludes with a discussion of the issues for consideration and recommendations for inclusion of SNF assemblies exposed to BPRs in criticality safety analyses using burnup credit for dry cask storage and transport.« less

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M

2015-01-01

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of latticemore » design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.« less

Modular fuel-cell stack assembly

DOEpatents

Patel, Pinakin [Danbury, CT; Urko, Willam [West Granby, CT

2008-01-29

A modular multi-stack fuel-cell assembly in which the fuel-cell stacks are situated within a containment structure and in which a gas distributor is provided in the structure and distributes received fuel and oxidant gases to the stacks and receives exhausted fuel and oxidant gas from the stacks so as to realize a desired gas flow distribution and gas pressure differential through the stacks. The gas distributor is centrally and symmetrically arranged relative to the stacks so that it itself promotes realization of the desired gas flow distribution and pressure differential.

Precise calculation of neutron-capture reactions contribution in energy release for different types of VVER-1000 fuel assemblies

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Bahdanovich, Rynat; Pham, Phu

2017-09-01

Precise calculation of energy release in a nuclear reactor is necessary to obtain the correct spatial power distribution and predict characteristics of burned nuclear fuel. In this work, previously developed method for calculation neutron-capture reactions - capture component - contribution in effective energy release in a fuel core of nuclear reactor is discussed. The method was improved and implemented to the different models of VVER-1000 reactor developed for MCU 5 and MCNP 4 computer codes. Different models of equivalent cell and fuel assembly in the beginning of fuel cycle were calculated. These models differ by the geometry, fuel enrichment and presence of burnable absorbers. It is shown, that capture component depends on fuel enrichment and presence of burnable absorbers. Its value varies for different types of hot fuel assemblies from 3.35% to 3.85% of effective energy release. Average capture component contribution in effective energy release for typical serial fresh fuel of VVER-1000 is 3.5%, which is 7 MeV/fission. The method will be used in future to estimate the dependency of capture energy on fuel density, burn-up, etc.

Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Banerjee, Kaushik; Clarity, Justin B; Cumberland, Riley M

This will be licensed via RSICC. A new, integrated data and analysis system has been designed to simplify and automate the performance of accurate and efficient evaluations for characterizing the input to the overall nuclear waste management system -UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database within UNF-ST&DARDS provides a standard means by which UNF-ST&DARDS can succinctly store and retrieve modeling and simulation (M&S) parameters for specific spent nuclear fuel analysis. A library of various analysis model templates provides the ability to communicate the various set of M&S parameters to the most appropriate M&S application.more » Interactive visualization capabilities facilitate data analysis and results interpretation. UNF-ST&DARDS current analysis capabilities include (1) assembly-specific depletion and decay, (2) and spent nuclear fuel cask-specific criticality and shielding. Currently, UNF-ST&DARDS uses SCALE nuclear analysis code system for performing nuclear analysis.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE PAGES

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; ...

2016-10-01

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

General view of the Space Shuttle Main Engine (SSME) assembly ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

General view of the Space Shuttle Main Engine (SSME) assembly with the expansion nozzle removed and resting on a cushioned mat on the floor of the SSME Processing Facility. The most prominent features in this view are the Low-pressure oxidizer Turbopump discharge Duct looping from the upper left side of the engine assembly to the lower left side of the assembly, the Low-Pressure Fuel Turbopump (LPFTP) is on the upper left of the assembly in this view and the LPFTP Discharge Duct loops from the upper left to upper right then turns back and down the assembly to the High-Pressure Fuel Turbopump on the lower right of the assembly. The Engine Controller and the Main fuel Valve Hydraulic Actuator are on the lower left portion of the assembly. The vertical rod that is in the approximate center of the engine assembly is a piece of ground support equipment call a Gimbal Actuator Replacement Strut which are used on the SSMEs when they are not installed in an orbiter. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

PBF Reactor Building (PER620). Fuel rod test assembly is on ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Fuel rod test assembly is on display at PBF. Date: 1982. INEEL negative no. 82-4893 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Membrane electrode assembly for a fuel cell

NASA Technical Reports Server (NTRS)

Prakash, Surya (Inventor); Narayanan, Sekharipuram R. (Inventor); Atti, Anthony (Inventor); Olah, George (Inventor); Smart, Marshall C. (Inventor)

2006-01-01

A catalyst ink for a fuel cell including a catalytic material and poly(vinylidene fluoride). The ink may be applied to a substrate to form an electrode, or bonded with other electrode layers to form a membrane electrode assembly (MEA).

Method of bonding

DOEpatents

Saller, deceased, Henry A.; Hodge, Edwin S.; Paprocki, Stanley J.; Dayton, Russell W.

1987-12-01

1. A method of making a fuel-containing structure for nuclear reactors, comprising providing an assembly comprising a plurality of fuel units; each fuel unit consisting of a core plate containing thermal-neutron-fissionable material, sheets of cladding metal on its bottom and top surfaces, said cladding sheets being of greater width and length than said core plates whereby recesses are formed at the ends and sides of said core plate, and end pieces and first side pieces of cladding metal of the same thickness as the core plate positioned in said recesses, the assembly further comprising a plurality of second side pieces of cladding metal engaging the cladding sheets so as to space the fuel units from one another, and a plurality of filler plates of an acid-dissolvable nonresilient material whose melting point is above 2000.degree. F., each filler plate being arranged between a pair of said second side pieces and the cladding plates of two adjacent fuel units, the filler plates having the same thickness as the second side pieces; the method further comprising enclosing the entire assembly in an envelope; evacuating the interior of the entire assembly through said envelope; applying inert gas under a pressure of about 10,000 psi to the outside of said envelope while at the same time heating the assembly to a temperature above the flow point of the cladding metal but below the melting point of any material of the assembly, whereby the envelope is pressed against the assembly and integral bonds are formed between plates, sheets, first side pieces, and end pieces and between the sheets and the second side pieces; slowly cooling the assembly to room temperature; removing the envelope; and dissolving the filler plates without attacking the cladding metal.

Storage, transportation and disposal system for used nuclear fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scaglione, John M.; Wagner, John C.

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

Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

NASA Astrophysics Data System (ADS)

Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

2001-02-01

Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented. .

Fuel burner and combustor assembly for a gas turbine engine

DOEpatents

Leto, Anthony

1983-01-01

A fuel burner and combustor assembly for a gas turbine engine has a housing within the casing of the gas turbine engine which housing defines a combustion chamber and at least one fuel burner secured to one end of the housing and extending into the combustion chamber. The other end of the fuel burner is arranged to slidably engage a fuel inlet connector extending radially inwardly from the engine casing so that fuel is supplied, from a source thereof, to the fuel burner. The fuel inlet connector and fuel burner coact to anchor the housing against axial movement relative to the engine casing while allowing relative radial movement between the engine casing and the fuel burner and, at the same time, providing fuel flow to the fuel burner. For dual fuel capability, a fuel injector is provided in said fuel burner with a flexible fuel supply pipe so that the fuel injector and fuel burner form a unitary structure which moves with the fuel burner.

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Favalli, Andrea; Vo, D.; Grogan, Brandon R.

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE PAGES

Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

2016-02-26

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Temporal switching of an amphiphilic self-assembly by a chemical fuel-driven conformational response† †Electronic supplementary information (ESI) available: Experimental section, synthetic procedures and supporting figures. See DOI: 10.1039/c7sc01730h Click here for additional data file.

PubMed Central

Jalani, Krishnendu; Dhiman, Shikha

2017-01-01

The spatial and temporal control of self-assemblies is the latest scientific hurdle in supramolecular chemistry which is inspired by the functioning of biological systems fueled by chemical signals. In this study, we work towards alleviating this scenario by employing a unique amphiphilic foldamer that operates under the effect of a chemical fuel. The conformational changes in the foldamer amplify into observable morphological changes in its amphiphilic assembly that are controlled by external molecular cues (fuel). We take advantage of this redox responsive foldamer to affect its conformation in a temporal manner by an enzymatic pathway. The temporal characteristics of the transient conformation/assembly can be modulated by varying the concentrations of the fuel and enzyme. We believe that such a design strategy can have positive consequences in designing molecular and supramolecular systems for future active, adaptive and autonomous materials. PMID:28989632

Upgrading the fuel-handling machine of the Novovoronezh nuclear power plant unit no. 5

NASA Astrophysics Data System (ADS)

Terekhov, D. V.; Dunaev, V. I.

2014-02-01

The calculation of safety parameters was carried out in the process of upgrading the fuel-handling machine (FHM) of the Novovoronezh nuclear power plant (NPP) unit no. 5 based on the results of quantitative safety analysis of nuclear fuel transfer operations using a dynamic logical-and-probabilistic model of the processing procedure. Specific engineering and design concepts that made it possible to reduce the probability of damaging the fuel assemblies (FAs) when performing various technological operations by an order of magnitude and introduce more flexible algorithms into the modernized FHM control system were developed. The results of pilot operation during two refueling campaigns prove that the total reactor shutdown time is lowered.

High-temperature, high-pressure optical port for rocket engine applications

NASA Technical Reports Server (NTRS)

Delcher, Ray; Nemeth, ED; Powers, W. T.

1993-01-01

This paper discusses the design, fabrication, and test of a window assembly for instrumentation of liquid-fueled rocket engine hot gas systems. The window was designed to allow optical measurements of hot gas in the SSME fuel preburner and appears to be the first window designed for application in a rocket engine hot gas system. Such a window could allow the use of a number of remote optical measurement technologies including: Raman temperature and species concentration measurement, Raleigh temperature measurements, flame emission monitoring, flow mapping, laser-induced florescence, and hardware imaging during engine operation. The window assembly has been successfully tested to 8,000 psi at 1000 F and over 11,000 psi at room temperature. A computer stress analysis shows the window will withstand high temperature and cryogenic thermal shock.

Direct methanol feed fuel cell with reduced catalyst loading

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor)

1999-01-01

Improvements to direct feed methanol fuel cells include new protocols for component formation. Catalyst-water repellent material is applied in formation of electrodes and sintered before application of ionomer. A membrane used in formation of an electrode assembly is specially pre-treated to improve bonding between catalyst and membrane. The improved electrode and the pre-treated membrane are assembled into a membrane electrode assembly.

Analysis of typical WWER-1000 severe accident scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sorokin, Yu.S.; Shchekoldin, V.V.; Borisov, L.N.

2004-07-01

At present in EDO 'Gidropress' there is a certain experience of performing the analyses of severe accidents of reactor plant with WWER with application of domestic and foreign codes. Important data were also obtained by the results of calculation modeling of integrated experiments with fuel assembly melting comprising a real fuel. Systematization and consideration of these data in development and assimilation of codes are extremely important in connection with large uncertainty still existing in understanding and adequate description of phenomenology of severe accidents. The presented report gives a comparison of analysis results of severe accidents of reactor plant with WWER-1000more » for two typical scenarios made by using American MELCOR code and the Russian RATEG/SVECHA/HEFEST code. The results of calculation modeling are compared using above codes with the data of experiment FPT1 with fuel assembly melting comprising a real fuel, which has been carried out at the facility Phebus (France). The obtained results are considered in the report from the viewpoint of: - adequacy of results of calculation modeling of separate phenomena during severe accidents of RP with WWER by using the above codes; - influence of uncertainties (degree of details of calculation models, choice of parameters of models etc.); - choice of those or other setup variables (options) in the used codes; - necessity of detailed modeling of processes and phenomena as applied to design justification of safety of RP with WWER. (authors)« less

Co-flow anode/cathode supply heat exchanger for a solid-oxide fuel cell assembly

DOEpatents

Haltiner, Jr., Karl J.; Kelly, Sean M.

2005-11-22

In a solid-oxide fuel cell assembly, a co-flow heat exchanger is provided in the flow paths of the reformate gas and the cathode air ahead of the fuel cell stack, the reformate gas being on one side of the exchanger and the cathode air being on the other. The reformate gas is at a substantially higher temperature than is desired in the stack, and the cathode gas is substantially cooler than desired. In the co-flow heat exchanger, the temperatures of the reformate and cathode streams converge to nearly the same temperature at the outlet of the exchanger. Preferably, the heat exchanger is formed within an integrated component manifold (ICM) for a solid-oxide fuel cell assembly.

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope; R. Sonat Sen; Brian Boer

2011-09-01

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code tomore » assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.« less

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

Numerical determination of lateral loss coefficients for subchannel analysis in nuclear fuel bundles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sin Kim; Goon-Cherl Park

1995-09-01

An accurate prediction of cross-flow based on detailed knowledge of the velocity field in subchannels of a nuclear fuel assembly is of importance in nuclear fuel performance analysis. In this study, the low-Reynolds number {kappa}-{epsilon} turbulence model has been adopted in two adjacent subchannels with cross-flow. The secondary flow is estimated accurately by the anisotropic algebraic Reynolds stress model. This model was numerically calculated by the finite element method and has been verified successfully through comparison with existing experimental data. Finally, with the numerical analysis of the velocity field in such subchannel domain, an analytical correlation of the lateral lossmore » coefficient is obtained to predict the cross-flow rate in subchannel analysis codes. The correlation is expressed as a function of the ratio of the lateral flow velocity to the donor subchannel axial velocity, recipient channel Reynolds number and pitch-to-diameter.« less

Compliant fuel cell system

DOEpatents

Bourgeois, Richard Scott [Albany, NY; Gudlavalleti, Sauri [Albany, NY

2009-12-15

A fuel cell assembly comprising at least one metallic component, at least one ceramic component and a structure disposed between the metallic component and the ceramic component. The structure is configured to have a lower stiffness compared to at least one of the metallic component and the ceramic component, to accommodate a difference in strain between the metallic component and the ceramic component of the fuel cell assembly.

PBF Reactor Building (PER620). PBF crane holds fuel test assembly ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). PBF crane holds fuel test assembly aloft prior to lowering into reactor for test. Date: 1982. INEEL negative no. 82-4909 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garner, P. L.; Hanan, N. A.

The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decidemore » to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.« less

System for fuel rod removal from a reactor module

DOEpatents

Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

1988-07-28

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

System for fuel rod removal from a reactor module

DOEpatents

Matchett, Richard L.; Roof, David R.; Kikta, Thomas J.; Wilczynski, Rosemarie; Nilsen, Roy J.; Bacvinskas, William S.; Fodor, George

1990-01-01

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

Pulverized fuel-oxygen burner

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor, Curtis; Patterson, Brad; Perdue, Jayson

A burner assembly combines oxygen and fuel to produce a flame. The burner assembly includes an oxygen supply tube adapted to receive a stream of oxygen and a solid fuel conduit arranged to extend through the oxygen tube to convey a stream of fluidized, pulverized, solid fuel into a flame chamber. Oxygen flowing through the oxygen supply tube passes generally tangentially through a first set of oxygen-injection holes formed in the solid fuel conduit and off-tangentially from a second set of oxygen-injection holes formed in the solid fuel conduit and then mixes with fluidized, pulverized, solid fuel passing through themore » solid fuel conduit to create an oxygen-fuel mixture in a downstream portion of the solid fuel conduit. This mixture is discharged into a flame chamber and ignited in the flame chamber to produce a flame.« less

A NEUTRONIC REACTOR

DOEpatents

Luebke, E.A.; Vandenberg, L.B.

1959-09-01

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1979-10-01

The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminary design for a high-burnup fuel rod, physics analyses of extended-burnup fuel cycles, studies of the physics characteristics of changes in fuel assembly metal-to-water ratios, and developmentmore » of a design concept for post-irradiation examination equipment to be utilized in examining high-burnup lead-test assemblies.« less

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1992-05-12

This patent describes a boiling water reactor having discrete bundles of fuel rods confined within channel enclosed fuel assemblies, an improvement to a fuel bundle assembly for placement in the reactor. It comprises a fuel channel having vertically extending walls forming a continuous channel around a fuel assembly volume, the channel being open at the bottom end for engagement to a lower tie plate and open at the upper end for engagement to an upper tie plate; rods for placement within the chamber, each the rod containing fissile material for producing nuclear reaction when in the presence of sufficient moderatedmore » neutron flux; a lower tie plate for supporting the bundle of rods within the channel, the lower tie plate for supporting the bundle of rods within the channel, the lower tie plate joining the bottom of the channel to close the bottom end of the channel, the lower tie plate providing defined apertures for the inflow of water in the channel between the rods for the generating of steam during the nuclear reaction; the plurality of fuel rods extending from the lower tie plate wherein a single phase region of the water in the bundle is defined to an upward portion of the bundle wherein a two phase region of the water and steam in the bundle is defined during nuclear steam generating reaction in the fuel bundle.« less

Air Breathing Direct Methanol Fuel Cell

DOEpatents

Ren; Xiaoming

2003-07-22

A method for activating a membrane electrode assembly for a direct methanol fuel cell is disclosed. The method comprises operating the fuel cell with humidified hydrogen as the fuel followed by running the fuel cell with methanol as the fuel.

Method and apparatus for selectively controlling the speed of an engine

DOEpatents

Davis, Roy Inge

2001-02-27

A control assembly 12 for use within a vehicle 10 having an engine 14 and which selectively controls the speed of the engine 14 in order to increase fuel efficiency and to effect relatively smooth starting and stopping of the engine. Particularly, in one embodiment, control assembly 12 cooperatively operates with a starter/alternator assembly 20 and is adapted for use with hybrid vehicles employing a start/stop powertrain assembly, wherein fuel efficiency is increased by selectively stopping engine operation when the vehicle has stopped.

Simulation numerique de l'effet du reflecteur radial sur les cellules rep en utilisant les codes DRAGON et DONJON

NASA Astrophysics Data System (ADS)

Bejaoui, Najoua

The pressurized water nuclear reactors (PWRs) is the largest fleet of nuclear reactors in operation around the world. Although these reactors have been studied extensively by designers and operators using efficient numerical methods, there are still some calculation weaknesses, given the geometric complexity of the core, still unresolved such as the analysis of the neutron flux's behavior at the core-reflector interface. The standard calculation scheme is a two steps process. In the first step, a detailed calculation at the assembly level with reflective boundary conditions, provides homogenized cross-sections for the assemblies, condensed to a reduced number of groups; this step is called the lattice calculation. The second step uses homogenized properties in each assemblies to calculate reactor properties at the core level. This step is called the full-core calculation or whole-core calculation. This decoupling of the two calculation steps is the origin of methodological bias particularly at the interface core reflector: the periodicity hypothesis used to calculate cross section librairies becomes less pertinent for assemblies that are adjacent to the reflector generally represented by these two models: thus the introduction of equivalent reflector or albedo matrices. The reflector helps to slowdown neutrons leaving the reactor and returning them to the core. This effect leads to two fission peaks in fuel assemblies localised at the core/reflector interface, the fission rate increasing due to the greater proportion of reentrant neutrons. This change in the neutron spectrum arises deep inside the fuel located on the outskirts of the core. To remedy this we simulated a peripheral assembly reflected with TMI-PWR reflector and developed an advanced calculation scheme that takes into account the environment of the peripheral assemblies and generate equivalent neutronic properties for the reflector. This scheme is tested on a core without control mechanisms and charged with fresh fuel. The results of this study showed that explicit representation of reflector and calculation of peripheral assembly with our advanced scheme allow corrections to the energy spectrum at the core interface and increase the peripheral power by up to 12% compared with that of the reference scheme.

CHF considerations for highly moderated 100% MOX fuels PWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saphier, D.; Raymond, P.

1995-09-01

A feasibility study on using 100% MOX fuel in a PWR with increased moderating ratio, RMA, was initiated. In the proposed design all the parameters were chosen identical to the French 1450MW PWR, except the fuel pin diameter which was reduced to achieve higher moderating ratios, V{sub M}/V{sub F}, where V{sub M} and V{sub F} are the moderator and fuel volume respectively. Moderating ratios from 2 to 4 were considered. In the present study the thermal-hydraulic feasibility of using fuel assemblies with smaller diameter fuel pins was investigated. The major design constrain in this study was the critical heat fluxmore » (CHF). In order to maintain the fuel pin integrity under nominal operating and transient conditions, the minimum DNBR, (Departure from Nucleate Boiling Ratio given by CHF/q{close_quotes}{sub local}, where q{close_quotes}{sub local} is the local heat flux), has to be above a given value. The limitations of the existing CHF correlations for the present study are outlined. Two designs based on the conventional 17x17 fuel assembly and on the advanced 19x19 assembly meeting the MDNBR criteria and satisfying the control margin requirements, are proposed.« less

Fuel-air munition and device

DOEpatents

Carlson, Gary A.

1976-01-01

An aerially delivered fuel-air munition consisting of an impermeable tank filled with a pressurized liquid fuel and joined at its two opposite ends with a nose section and a tail assembly respectively to complete an aerodynamic shape. On impact the tank is explosively ruptured to permit dispersal of the fuel in the form of a fuel-air cloud which is detonated after a preselected time delay by means of high explosive initiators ejected from the tail assembly. The primary component in the fuel is methylacetylene, propadiene, or mixtures thereof to which is added a small mole fraction of a relatively high vapor pressure liquid diluent or a dissolved gas diluent having a low solubility in the primary component.

Fuel cell with electrolyte feed system

DOEpatents

Feigenbaum, Haim

1984-01-01

A fuel cell having a pair of electrodes at the sites of electrochemical reactions of hydrogen and oxygen and a phosphoric acid electrolyte provided with an electrolyte supporting structure in the form of a laminated matrix assembly disposed between the electrodes. The matrix assembly is formed of a central layer disposed between two outer layers, each being permeable to the flow of the electrolyte. The central layer is provided with relatively large pores while the outer layers are provided with relatively small pores. An external reservoir supplies electrolyte via a feed means to the central layer to compensate for changes in electrolyte volume in the matrix assembly during the operation of fuel cell.

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

DOE PAGES

Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.; ...

2017-10-07

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

NASA Astrophysics Data System (ADS)

Menlove, Howard; Belian, Anthony; Geist, William; Rael, Carlos

2018-01-01

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd2O3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate the 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. We have named the system the fast-neutron passive collar (FNPC).

Fuel cell system with separating structure bonded to electrolyte

DOEpatents

Bourgeois, Richard Scott; Gudlavalleti, Sauri; Quek, Shu Ching; Hasz, Wayne Charles; Powers, James Daniel

2010-09-28

A fuel cell assembly comprises a separating structure configured for separating a first reactant and a second reactant wherein the separating structure has an opening therein. The fuel cell assembly further comprises a fuel cell comprising a first electrode, a second electrode, and an electrolyte interposed between the first and second electrodes, and a passage configured to introduce the second reactant to the second electrode. The electrolyte is bonded to the separating structure with the first electrode being situated within the opening, and the second electrode being situated within the passage.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

SFCOMPO 2.0 - A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

NASA Astrophysics Data System (ADS)

Michel-Sendis, Franco; Martinez-González, Jesus; Gauld, Ian

2017-09-01

SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF) samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA). The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

Integral edge seals for phosphoric acid fuel cells

DOEpatents

Granata, Jr., Samuel J.; Woodle, Boyd M.; Dunyak, Thomas J.

1992-01-01

A phosphoric acid fuel cell having integral edge seals formed by an elastomer permeating an outer peripheral band contiguous with the outer peripheral edges of the cathode and anode assemblies and the matrix to form an integral edge seal which is reliable, easy to manufacture and has creep characteristics similar to the anode, cathode and matrix assemblies inboard of the seals to assure good electrical contact throughout the life of the fuel cell.

Vibrating fuel grapple

DOEpatents

Chertock, deceased, Alan J.; Fox, Jack N.; Weissinger, Robert B.

1982-01-01

A reactor refueling method utilizing a vibrating fuel grapple for removing spent fuel assemblies from a reactor core which incorporates a pneumatic vibrator in the grapple head, enabling additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

Measuring the Multiplication of Spent Fuel Assemblies – It’s easier than you think!

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph

This is a set of eight slides which advertise how easy it can be to measure the multiplication of a spent fuel assembly. A robust (fission chambers), rapid (under 15 minutes), direct (multiplication is measured, not photons from fission fragments) measurement of multiplication is possible.

Validation of the U.S. NRC NGNP evaluation model with the HTTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saller, T.; Seker, V.; Downar, T.

2012-07-01

The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water reactor (LWR) homogenization methods rely on the short mean free paths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTR neutrons have much longer mean free paths and penetrate further into neighboring assemblies than in LWRs. Because of this, conventional lattice calculations with a single assembly may not be valid. In addition to difficulties caused by the longer mean free paths, the HTTR presents unique axial and radial heterogeneities that require additional modifications to the single assembly homogenization method. To handle these challenges, the homogenizationmore » domain is decreased while the computational domain is increased. Instead of homogenizing a single hexagonal fuel assembly, the assembly is split into six triangles on the radial plane and five blocks axially in order to account for the placement of burnable poisons. Furthermore, the radial domain is increased beyond a single fuel assembly to account for spectrum effects from neighboring fuel, reflector, and control rod assemblies. A series of five two-dimensional cases, each closer to the full core, were calculated to evaluate the effectiveness of the homogenization method and cross-sections. (authors)« less

Development and applications of retro-reflective surfaces for ultrasound in LBE

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

The Belgian Nuclear Research Centre SCK-CEN is in the process of developing MYRRHA - a fast flux research reactor to replace the aging BR2. MYRRHA is conceptualized as an accelerator driven system cooled by lead bismuth eutectic mixture (LBE). As LBE is opaque to visual light, ultrasonic measurement techniques are employed as the main technology to provide feedback to submerged operations when needed. Conceptually, MYRRHA is a pool type reactor divided in a cold lower plenum and hot upper plenum separated by a diaphragm that forces the main flow through the core. The main flow is cooled by four heatmore » exchangers and driven by two liquid metal pumps. One of the tasks tackled using ultrasound is locating a potentially lost fuel assembly to assist a recovery operation. As all fuel manipulations in MYRRHA are performed in the lower plenum, a potentially lost fuel assembly is located in the lower plenum. Buoyancy will force the lost fuel assembly to float against the diaphragm unless it is still partially inserted in the core. Because of the latter situation, an ultrasonic scan localizing the fuel assembly should be performed from a large distance to avoid a collision with such a partially inserted fuel assembly. Unfortunately, standard machined stainless steel objects, such as a fuel assembly, reflect an ultrasonic pulse in a specular way which induces stringent requirements on the alignment of the ultrasonic sensor with respect to the fuel assembly as we cannot rely on diffuse reflections and/or scattering of the ultrasonic pulse. Moreover, increasing the distance also increases geometric spreading and absorption of the pulse weakening the signal amplitude even faster to noise levels when deviating from perfect alignment. An alternative approach consists in relying on reflections from the known surroundings: a lost fuel assembly will block the line-of-sight to the diaphragm resulting in an anomaly in the reflection - either a shorter than expected time-of-flight of the pulse or a complete absence of a reflection like a shadow. In that way, it suffices to align the sensor with the diaphragm instead of the fuel assembly which is much easier to achieve as the robotics on which the sensor is mounted move parallel with the diaphragm. The alignment requirement in the latter approach can be further relaxed by using a tiling of retro-reflectors on the lower surface of the diaphragm. In that way, alignment becomes less vital and the main source of acoustic energy loss - geometric spread of the beam - is almost completely removed, leaving only absorption losses. In this paper, we present the first results in developing a retro reflectance surface for ultrasound in LBE. We present experimental results for different designs of retro-reflectors in both water and LBE. We discuss both linear and array retro-reflectors of different sizes and investigate the influence of the main relevant ultrasonic parameters such as wavelength and spot size on the strength of the received reflection under different alignment angles. We also demonstrate how retro-reflective surfaces can be exploited when localizing objects using linear and rotating scanning methods. (authors)« less

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

Lead Slowing-Down Spectrometry Time Spectral Analysis for Spent Fuel Assay: FY11 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulisek, Jonathan A.; Anderson, Kevin K.; Bowyer, Sonya M.

2011-09-30

Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration, of which PNNL is a part, to study the feasibility of Lead Slowing Down Spectroscopy (LSDS). This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertaintymore » considerably lower than the approximately 10% typical of today's confirmatory assay methods. This document is a progress report for FY2011 PNNL analysis and algorithm development. Progress made by PNNL in FY2011 continues to indicate the promise of LSDS analysis and algorithms applied to used fuel. PNNL developed an empirical model based on calibration of the LSDS to responses generated from well-characterized used fuel. The empirical model, which accounts for self-shielding effects using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the true self-shielding functions of the used fuel assembly models. The potential for the direct and independent assay of the sum of the masses of 239Pu and 241Pu to within approximately 3% over a wide used fuel parameter space was demonstrated. Also, in FY2011, PNNL continued to develop an analytical model. Such efforts included the addition of six more non-fissile absorbers in the analytical shielding function and the non-uniformity of the neutron flux across the LSDS assay chamber. A hybrid analytical-empirical approach was developed to determine the mass of total Pu (sum of the masses of 239Pu, 240Pu, and 241Pu), which is an important quantity in safeguards. Results using this hybrid method were of approximately the same accuracy as the pure empirical approach. In addition, total Pu with much better accuracy with the hybrid approach than the pure analytical approach. In FY2012, PNNL will continue efforts to optimize its empirical model and minimize its reliance on calibration data. In addition, PNNL will continue to develop an analytical model, considering effects such as neutron-scattering in the fuel and cladding, as well as neutrons streaming through gaps between fuel pins in the fuel assembly.« less

Sensitivity and Uncertainty Analysis of the GFR MOX Fuel Subassembly

NASA Astrophysics Data System (ADS)

Lüley, J.; Vrban, B.; Čerba, Š.; Haščík, J.; Nečas, V.; Pelloni, S.

2014-04-01

We performed sensitivity and uncertainty analysis as well as benchmark similarity assessment of the MOX fuel subassembly designed for the Gas-Cooled Fast Reactor (GFR) as a representative material of the core. Material composition was defined for each assembly ring separately allowing us to decompose the sensitivities not only for isotopes and reactions but also for spatial regions. This approach was confirmed by direct perturbation calculations for chosen materials and isotopes. Similarity assessment identified only ten partly comparable benchmark experiments that can be utilized in the field of GFR development. Based on the determined uncertainties, we also identified main contributors to the calculation bias.

Fibre optic sensor based measurements of flow-induced vibration in a liquid metal cooled nuclear reactor set-up

NASA Astrophysics Data System (ADS)

De Pauw, B.; Vanlanduit, S.; Van Tichelen, K.; Geernaert, T.; Thienpont, H.; Berghmans, F.

2017-04-01

Fuel assembly vibrations in nuclear reactor cores should not be excessive as these can compromise the lifetime of the assembly and lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. We therefore demonstrate accurate measurements of the vibrations of a fuel assembly in a lead-bismuth eutectic cooled installation with fibre Bragg grating (FBG) based sensors. The use of FBGs in combination with a dedicated sensor integration approach allows accounting for the severe geometrical constraints and providing for the required minimal intrusiveness of the instrumentation, identifying the vibration modes with required accuracy and observing the differences between the vibration amplitudes of the individual fuel pins as well as evidencing a low frequency fuel pin vibration mode resulting from the supports.

Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach.

PubMed

Rochman, D; Vasiliev, A; Ferroukhi, H; Pecchia, M

2018-06-15

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS 2 M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO 2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources. Copyright © 2018 Elsevier B.V. All rights reserved.

Closeup view of the bottom area of Space Shuttle Main ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

Close-up view of the bottom area of Space Shuttle Main Engine (SSME) 2052 engine assembly mounted in a SSME Engine Handler in the Horizontal Processing area of the SSME Processing Facility at Kennedy Space Center. The most prominent features in this view are the Low-Pressure Oxidizer Discharge Duct toward the bottom of the assembly, the SSME Engine Controller and the Main Fuel Valve Hydraulic Actuator are in the approximate center of the assembly in this view, the Low-Pressure Fuel Turbopump (LPFTP), the LPFTP Discharge Duct are to the left on the assembly in this view and the High-Pressure Fuel Turbopump is located toward the top of the engine assembly in this view. The ring of tabs in the right side of the image, at the approximate location of the Nozzle and the Coolant Outlet Manifold interface is the Heat Shield Support Ring. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

NASA Technical Reports Server (NTRS)

Rohal, R. G.; Tambling, T. N.; Smith, R. L.

1973-01-01

Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

Gamma and fast neutron radiation monitoring inside spent reactor fuel assemblies

NASA Astrophysics Data System (ADS)

Lakosi, L.; Tam Nguyen, C.

2007-09-01

Gamma and neutron signatures of spent reactor fuel were monitored by small-size silicon diode and track etch detectors, respectively, in a nuclear power plant (NPP). These signatures, reflecting gross gamma intensity and the 242,244Cm content, contain information on the burn-up (BU) and cooling time (CT) of the fuel. The small size of the detectors allows close access to inside parts of the assemblies out of reach of other methods. A commercial Si diode was encapsulated in a cylindrical steel case and was used for gross γ monitoring. CR-39 detectors were used for neutron measurements. Irradiation exposures at the NPP were implemented in the central dosimetric channel of spent fuel assemblies (SFAs) stored in borated water. Gross γ and neutron axial profiles were taken up by scanning with the aid of a long steel guide tube, lowered down to the spent fuel pond by crane and fitted to the headpiece of the fuel assemblies. Gamma measurements were performed using a long cable introduced in this tube, with the Si diode at the end. A long steel wire was also led through the guide tube, to which a chain of 15 sample holder capsules was attached, each containing a track detector. Gamma dose rates of 0.1-10 kGy h -1, while neutron fluxes in a range of (0.25-26) 10 4 cm -2 s -1 were recorded. The results are in good correlation with those of a calculation for spent fuel neutron yield.

Saturn Apollo Program

NASA Image and Video Library

1964-09-01

This photograph shows the components for the Saturn V S-IC stage fuel tank assembly in the Manufacturing Engineering Laboratory, building 4707, at the Marshall Space Flight Center (MSFC). Left to right are upper head, lower head, and forward skirt assembly. Thirty-three feet in diameter, they will hold a total of 4,400,000 pounds of fuel. Although this tankage was assembled at MSFC, the elements were made by the Boeing Company at Wichita and the Michould Operations at New Orleans.

Cooling assembly for fuel cells

DOEpatents

Kaufman, Arthur; Werth, John

1990-01-01

A cooling assembly for fuel cells having a simplified construction whereby coolant is efficiently circulated through a conduit arranged in serpentine fashion in a channel within a member of such assembly. The channel is adapted to cradle a flexible, chemically inert, conformable conduit capable of manipulation into a variety of cooling patterns without crimping or otherwise restricting of coolant flow. The conduit, when assembled with the member, conforms into intimate contact with the member for good thermal conductivity. The conduit is non-corrodible and can be constructed as a single, manifold-free, continuous coolant passage means having only one inlet and one outlet.

System Regulates the Water Contents of Fuel-Cell Streams

NASA Technical Reports Server (NTRS)

Vasquez, Arturo; Lazaroff, Scott

2005-01-01

An assembly of devices provides for both humidification of the reactant gas streams of a fuel cell and removal of the product water (the water generated by operation of the fuel cell). The assembly includes externally-sensing forward-pressure regulators that supply reactant gases (fuel and oxygen) at variable pressures to ejector reactant pumps. The ejector supply pressures depend on the consumption flows. The ejectors develop differential pressures approximately proportional to the consumption flow rates at constant system pressure and with constant flow restriction between the mixer-outlet and suction ports of the ejectors. For removal of product water from the circulating oxygen stream, the assembly includes a water/gas separator that contains hydrophobic and hydrophilic membranes. The water separator imposes an approximately constant flow restriction, regardless of the quality of the two-phase flow that enters it from the fuel cell. The gas leaving the water separator is nearly 100 percent humid. This gas is returned to the inlet of the fuel cell along with a quantity of dry incoming oxygen, via the oxygen ejector, thereby providing some humidification.

Apparatus for inspecting fuel elements

DOEpatents

Oakley, David J.; Groves, Oliver J.; Kaiser, Bruce J.

1986-01-01

Disclosed is an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

Apparatus for inspecting fuel elements

DOEpatents

Kaiser, B.J.; Oakley, D.J.; Groves, O.J.

1984-12-21

This disclosure describes an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

Electrolyte creepage barrier for liquid electrolyte fuel cells

DOEpatents

Li, Jian [Alberta, CA; Farooque, Mohammad [Danbury, CT; Yuh, Chao-Yi [New Milford, CT

2008-01-22

A dielectric assembly for electrically insulating a manifold or other component from a liquid electrolyte fuel cell stack wherein the dielectric assembly includes a substantially impermeable dielectric member over which electrolyte is able to flow and a barrier adjacent the dielectric member and having a porosity of less than 50% and greater than 10% so that the barrier is able to measurably absorb and chemically react with the liquid electrolyte flowing on the dielectric member to form solid products which are stable in the liquid electrolyte. In this way, the barrier inhibits flow or creepage of electrolyte from the dielectric member to the manifold or component to be electrically insulated from the fuel cell stack by the dielectric assembly.

Fuel cell assembly with electrolyte transport

DOEpatents

Chi, Chang V.

1983-01-01

A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

MELCOR model for an experimental 17x17 spent fuel PWR assembly.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey

2010-11-01

A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru; Pinegin, A. A.

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

Vibrating fuel grapple. [LMFBR

DOEpatents

Chertock, A.J.; Fox, J.N.; Weissinger, R.B.

A reactor refueling method is described which utilizes a vibrating fuel grapple for removing spent fuel assemblies from a reactor core. It incorporates a pneumatic vibrator in the grapple head which allows additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

TMI-2 (Three Mile Island Unit 2) core region defueling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodabaugh, J.M.; Cowser, D.K.

1988-01-01

In July of 1982, a video camera was inserted into the Three Mile Island Unit 2 reactor vessel providing the first visual evidence of core damage. This inspection, and numerous subsequent data acquisition tasks, revealed a central void /approx/1.5 m (5 ft) deep. This void region was surrounded by partial length fuel assemblies and ringed on the periphery by /approx/40 full-length, but partial cross-section, fuel assemblies. All of the original 177 fuel assemblies exhibited signs of damage. The bottom of the void cavity was covered with a bed of granular rubble, fuel assembly upper end fittings, control rod spiders, fuelmore » rod fragments, and fuel pellets. It was obvious that the normal plant refueling system not suitable for removing the damaged core. A new system of defueling tools and equipment was necessary to perform this task. Design of the new system was started immediately, followed by >1 yr of fabrication. Delivery and checkout of the defueling system occurred in mid-1985. Actual defueling was initiated in late 1985 with removal of the debris bed at the bottom of the core void. Obstructions to the debris, such as end fittings and fuel rod fragments ere removed first; then /approx/23,000 kg (50,000lb) of granular debris was quickly loaded into canisters. Core region defueling was completed in late 1987, /approx/2 yr after it was initiated.« less

Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

DOE Office of Scientific and Technical Information (OSTI.GOV)

Patra, Anirban; Tomé, Carlos N.

A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

DOE PAGES

Patra, Anirban; Tomé, Carlos N.

2017-03-06

A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, M. A.; Sen, R. S.; Ougouag, A. M.

2012-07-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space availablemore » for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)« less

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag

2012-04-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel,more » the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.« less

Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

NASA Astrophysics Data System (ADS)

Tucker, Lucas Powelson

This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

Chalcogen catalysts for polymer electrolyte fuel cell

DOEpatents

Alonso-Vante, Nicolas [Buxerolles, FR; Zelenay, Piotr [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Wieckowski, Andrzej [Champaign, IL; Cao, Dianxue [Urbana, IL

2009-09-15

A methanol-tolerant cathode catalyst and a membrane electrode assembly for fuel cells that includes such a cathode catalyst. The cathode catalyst includes a support having at least one transition metal in elemental form and a chalcogen disposed on the support. Methods of making the cathode catalyst and membrane electrode assembly are also described.

Mixer Assembly for a Gas Turbine Engine

NASA Technical Reports Server (NTRS)

Smith, Lance L. (Inventor); Fotache, Catalin G. (Inventor); Dai, Zhongtao (Inventor); Cohen, Jeffrey M. (Inventor); Hautman, Donald J. (Inventor)

2015-01-01

A mixer assembly for a gas turbine engine is provided, including a main mixer with fuel injection holes located between at least one radial swirler and at least one axial swirler, wherein the fuel injected into the main mixer is atomized and dispersed by the air flowing through the radial swirler and the axial swirler.

Mixer Assembly for a Gas Turbine Engine

NASA Technical Reports Server (NTRS)

Dai, Zhongtao (Inventor); Cohen, Jeffrey M. (Inventor); Fotache, Catalin G. (Inventor); Hautman, Donald J. (Inventor); Smith, Lance L. (Inventor)

2018-01-01

A mixer assembly for a gas turbine engine is provided, including a main mixer with fuel injection holes located between at least one radial swirler and at least one axial swirler, wherein the fuel injected into the main mixer is atomized and dispersed by the air flowing through the radial swirler and the axial swirler.

Chalcogen catalysts for polymer electrolyte fuel cell

DOEpatents

Zelenay, Piotr; Choi, Jong-Ho; Alonso-Vante, Nicolas; Wieckowski, Andrzej; Cao, Dianxue

2010-08-24

A methanol-tolerant cathode catalyst and a membrane electrode assembly for fuel cells that includes such a cathode catalyst. The cathode catalyst includes a support having at least one transition metal in elemental form and a chalcogen disposed on the support. Methods of making the cathode catalyst and membrane electrode assembly are also described.

Additively Manufactured Main Fuel Valve Housing

NASA Technical Reports Server (NTRS)

Eddleman, David; Richard, Jim

2015-01-01

Selective Laser Melting (SLM) was utilized to fabricate a liquid hydrogen valve housing typical of those found in rocket engines and main propulsion systems. The SLM process allowed for a valve geometry that would be difficult, if not impossible to fabricate by traditional means. Several valve bodies were built by different SLM suppliers and assembled with valve internals. The assemblies were then tested with liquid nitrogen and operated as desired. One unit was also burst tested and sectioned for materials analysis. The design, test results, and planned testing are presented herein.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

Combination of biodiesel-ethanol-diesel fuel blend and SCR catalyst assembly to reduce emissions from a heavy-duty diesel engine.

PubMed

Shi, Xiaoyan; Yu, Yunbo; He, Hong; Shuai, Shijin; Dong, Hongyi; Li, Rulong

2008-01-01

In this study, the efforts to reduce NOx and particulate matter (PM) emissions from a diesel engine using both ethanol-selective catalytic reduction (SCR) of NOx over an Ag/Al2O3 catalyst and a biodiesel-ethanol-diesel fuel blend (BE-diesel) on an engine bench test are discussed. Compared with diesel fuel, use of BE-diesel increased PM emissions by 14% due to the increase in the soluble organic fraction (SOF) of PM, but it greatly reduced the Bosch smoke number by 60%-80% according to the results from 13-mode test of European Stationary Cycle (ESC) test. The SCR catalyst was effective in NOx reduction by ethanol, and the NOx conversion was approximately 73%. Total hydrocarbons (THC) and CO emissions increased significantly during the SCR of NOx process. Two diesel oxidation catalyst (DOC) assemblies were used after Ag/Al2O3 converter to remove CO and HC. Different oxidation catalyst showed opposite effect on PM emission. The PM composition analysis revealed that the net effect of oxidation catalyst on total PM was an integrative effect on SOF reduction and sulfate formation of PM. The engine bench test results indicated that the combination of BE-diesel and a SCR catalyst assembly could provide benefits for NOx and PM emissions control even without using diesel particle filters (DPFs).

Method of making MEA for PEM/SPE fuel cell

DOEpatents

Hulett, Jay S.

2000-01-01

A method of making a membrane-electrode-assembly (MEA) for a PEM/SPE fuel cell comprising applying a slurry of electrode-forming material directly onto a membrane-electrolyte film. The slurry comprises a liquid vehicle carrying catalyst particles and a binder for the catalyst particles. The membrane-electrolyte is preswollen by contact with the vehicle before the electrode-forming slurry is applied to the membrane-electrolyte. The swollen membrane-electrolyte is constrained against shrinking in the "x" and "y" directions during drying. Following assembly of the fuel cell, the MEA is rehydrated inside the fuel cell such that it swells in the "z" direction for enhanced electrical contact with contiguous electrically conductive components of the fuel cell.

Detachable connection for a nuclear reactor fuel assembly

DOEpatents

Christiansen, D.W.; Karnesky, R.A.

1983-08-29

A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

Process for recycling components of a PEM fuel cell membrane electrode assembly

DOEpatents

Shore, Lawrence [Edison, NJ

2012-02-28

The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

Detachable connection for a nuclear reactor fuel assembly

DOEpatents

Christiansen, David W.; Karnesky, Richard A.

1986-01-01

A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

Measuring spent fuel assembly multiplication in borated water with a passive neutron albedo reactivity instrument

NASA Astrophysics Data System (ADS)

Tobin, Stephen J.; Peura, Pauli; Bélanger-Champagne, Camille; Moring, Mikael; Dendooven, Peter; Honkamaa, Tapani

2018-07-01

The performance of a passive neutron albedo reactivity (PNAR) instrument to measure neutron multiplication of spent nuclear fuel in borated water is investigated as part of an integrated non-destructive assay safeguards system. To measure the PNAR Ratio, which is proportional to the neutron multiplication, the total neutron count rate is measured in high- and low-multiplying environments by the PNAR instrument. The integrated system also contains a load cell and a passive gamma emission tomograph, and as such meets all the recommendations of the IAEA's recent ASTOR Experts Group report. A virtual spent fuel library for VVER-440 fuel was used in conjunction with MCNP simulations of the PNAR instrument to estimate the measurement uncertainties from (1) variation in the water boron content, (2) assembly positioning in the detector and (3) counting statistics. The estimated aggregate measurement uncertainty on the PNAR Ratio measurement is 0.008, to put this uncertainty in context, the difference in the PNAR Ratio between a fully irradiated assembly and this same assembly when fissile isotopes only absorb neutrons, but do not emit neutrons, is 0.106, a 13-sigma effect. The 1-sigma variation of 0.008 in the PNAR Ratio is estimated to correspond to a 3.2 GWd/tU change in assembly burnup.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, Charles W.

1998-01-01

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, C.W.

1998-11-03

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

NNSA B-Roll: MOX Facility

ScienceCinema

None

2017-12-09

In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

NNSA B-Roll: MOX Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2010-05-21

In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

All About MOX

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2009-07-29

In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

FUEL ASSEMBLY FOR A NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-29

A fuel assembly for a nuclear reactor of the type wherein liquid coolant is circulated through the core of the reactor in contact with the external surface of the fuel elements is described. In this design a plurality of parallel plates containing fissionable material are spaced about one-tenth of an inch apart and are supported between a pair of spaced parallel side members generally perpendicular to the plates. The plates all have a small continuous and equal curvature in the same direction between the side members.

All About MOX

ScienceCinema

None

2018-01-16

In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

Fabrication Method for Laboratory-Scale High-Performance Membrane Electrode Assemblies for Fuel Cells.

PubMed

Sassin, Megan B; Garsany, Yannick; Gould, Benjamin D; Swider-Lyons, Karen E

2017-01-03

Custom catalyst-coated membranes (CCMs) and membrane electrode assemblies (MEAs) are necessary for the evaluation of advanced electrocatalysts, gas diffusion media (GDM), ionomers, polymer electrolyte membranes (PEMs), and electrode structures designed for use in next-generation fuel cells, electrolyzers, or flow batteries. This Feature provides a reliable and reproducible fabrication protocol for laboratory scale (10 cm 2 ) fuel cells based on ultrasonic spray deposition of a standard Pt/carbon electrocatalyst directly onto a perfluorosulfonic acid PEM.

Fuel inspection and reconstitution experience at Surry Power Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brookmire, T.A.

Surry Power Station, located on the James River near Williamsburg, Virginia, has two Westinghouse pressurized water reactors. Unit 2 consistently sets a high standard of fuel performance (no indication of fuel failures in recent cycles), while unit 1, since cycle 6, has been plagued with numerous fuel failures. Both Surry units operate with Westinghouse standard 15 x 15 fuel. Virginia Power management set goals to reduce the coolant activity, thus reducing person-rem exposure and the associated costs of high coolant activity. To achieve this goal, extensive fuel examination campaigns were undertaken that included high-magnification video inspectionsa, debris cleaning, wet andmore » vacuum fuel sipping, fuel rod ultrasonic testing, and eddy current examination. In the summer of 1985, during cycle 8 operation, Kraftwerk Union reconstituted (repaired) the damage, once-burned assemblies from cycles 6 and 7 by replacing failed fuel rods with solid Zircaloy-4 rods. Currently, cycle 9 has operated for 5 months without any indication of fuel failure (the cycle 9 core has two reconstituted assemblies).« less

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

DOE PAGES

Carmack, W. Jon; Chichester, Heather M.; Porter, Douglas L.; ...

2016-02-27

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This then places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. After comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, W. J.; Chichester, H. M.; Porter, D. L.

2016-05-01

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huizenga, D. J.

2016-06-08

A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspectsmore » of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.« less

Proton-Fueled, Reversible DNA Hybridization Chain Assembly for pH Sensing and Imaging.

PubMed

Liu, Lan; Liu, Jin-Wen; Huang, Zhi-Mei; Wu, Han; Li, Na; Tang, Li-Juan; Jiang, Jian-Hui

2017-07-05

Design of DNA self-assembly with reversible responsiveness to external stimuli is of great interest for diverse applications. We for the first time develop a pH-responsive, fully reversible hybridization chain reaction (HCR) assembly that allows sensitive sensing and imaging of pH in living cells. Our design relies on the triplex forming sequences that form DNA triplex with toehold regions under acidic conditions and then induce a cascade of strand displacement and DNA assembly. The HCR assembly has shown dynamic responses in physiological pH ranges with excellent reversibility and demonstrated the potential for in vitro detection and live-cell imaging of pH. Moreover, this method affords HCR assemblies with highly localized fluorescence responses, offering advantages of improving sensitivity and better selectivity. The proton-fueled, reversible HCR assembly may provide a useful approach for pH-related cell biology study and disease diagnostics.

Integral gas seal for fuel cell gas distribution assemblies and method of fabrication

DOEpatents

Dettling, Charles J.; Terry, Peter L.

1985-03-19

A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

Method of fabricating an integral gas seal for fuel cell gas distribution assemblies

DOEpatents

Dettling, Charles J.; Terry, Peter L.

1988-03-22

A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

Grid-to-rod flow-induced impact study for PWR fuel in reactor

DOE PAGES

Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

2016-06-10

The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

NASA Astrophysics Data System (ADS)

Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

2018-01-01

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

Manufacturing and Performance Assessment of Stamped, Laser Welded, and Nitrided FeCrV Stainless Steel Bipolar Plates for Proton Exchange Membrane Fuel Cells

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brady, Michael P; Abdelhamid, Mahmoud; Dadheech, G

A manufacturing and single-cell fuel cell performance study of stamped, laser welded, and gas nitrided ferritic stainless steel foils in an advanced automotive bipolar plate assembly design was performed. Two developmental foil compositions were studied: Fee20Cre4V and Fee23Cre4V wt.%. Foils 0.1 mm thick were stamped and then laser welded together to create single bipolar plate assemblies with cooling channels. The plates were then surface treated by pre-oxidation and nitridation in N2e4H2 based gas mixtures using either a conventional furnace or a short-cycle quartz lamp infrared heating system. Single-cell fuel cell testing was performed at 80 C for 500 h atmore » 0.3 A/cm2 using 100% humidification and a 100%/40% humidification cycle that stresses the membrane and enhances release of the fluoride ion and promotes a more corrosive environment for the bipolar plates. Periodic high frequency resistance potential-current scans during the 500 h fuel cell test and posttest analysis of the membrane indicated no resistance increase of the plates and only trace levels of metal ion contamination.« less

Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dawson, D.M.; Guerra, G.; Neider, T.

1995-12-01

This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity ormore » other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.« less

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

DOE PAGES

Brown, C. S.; Zhang, Hongbin

2016-05-24

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

PBF Reactor Building (PER620). Detail of fuel test assembly in ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Towards neat methanol operation of direct methanol fuel cells: a novel self-assembled proton exchange membrane.

PubMed

Li, Jing; Cai, Weiwei; Ma, Liying; Zhang, Yunfeng; Chen, Zhangxian; Cheng, Hansong

2015-04-18

We report here a novel proton exchange membrane with remarkably high methanol-permeation resistivity and excellent proton conductivity enabled by carefully designed self-assembled ionic conductive channels. A direct methanol fuel cell utilizing the membrane performs well with a 20 M methanol solution, very close to the concentration of neat methanol.

Development and Application of a Sample Holder for In Situ Gaseous TEM Studies of Membrane Electrode Assemblies for Polymer Electrolyte Fuel Cells.

PubMed

Kamino, Takeo; Yaguchi, Toshie; Shimizu, Takahiro

2017-10-01

Polymer electrolyte fuel cells hold great potential for stationary and mobile applications due to high power density and low operating temperature. However, the structural changes during electrochemical reactions are not well understood. In this article, we detail the development of the sample holder equipped with gas injectors and electric conductors and its application to a membrane electrode assembly of a polymer electrolyte fuel cell. Hydrogen and oxygen gases were simultaneously sprayed on the surfaces of the anode and cathode catalysts of the membrane electrode assembly sample, respectively, and observation of the structural changes in the catalysts were simultaneously carried out along with measurement of the generated voltages.

Neutron Characterization of Encapsulated ATF-1/LANL-1 Mockup Fuel Capsules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vogel, Sven C.; Borges, Nicholas Paul; Losko, Adrian Simon

Twenty pellets of mock-up accident tolerant fuels UN-U3Si5 were produced at LANL and loaded in two rodlet/capsule assemblies. Tomographic imaging and diffraction measurements were performed to characterize these samples at the Flight-Path 5 and HIPPO beam lines at LANSCE/LANL between November 2016 and January 2017 as well as in August 2017. The entire ~10 cm long, ~1 cm diameter fuel volume could be characterized, however due to time constraints only 2 mm slices in 4mm increments were characterized with neutron diffraction and a 28mm subset of the entire sample was characterized with energy-resolved neutron imaging. The double encapsulation of themore » fuel into two steel containers does not pose a problem for the neutron analysis and the methods could be applied to enriched as well irradiated fuels.« less

Neutronics Investigations for the Lower Part of a Westinghouse SVEA-96+ Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, M.F.; Luethi, A.; Seiler, R.

2002-05-15

Accurate critical experiments have been performed for the validation of total fission (F{sub tot}) and {sup 238}U-capture (C{sub 8}) reaction rate distributions obtained with CASMO-4, HELIOS, BOXER, and MCNP4B for the lower axial region of a real Westinghouse SVEA-96+ fuel assembly. The assembly comprised fresh fuel with an average {sup 235}U enrichment of 4.02 wt%, a maximum enrichment of 4.74 wt%, 14 burnable-absorber fuel pins, and full-density water moderation. The experimental configuration investigated was core 1A of the LWR-PROTEUS Phase I project, where 61 different fuel pins, representing {approx}64% of the assembly, were gamma-scanned individually. Calculated (C) and measured (E)more » values have been compared in terms of C/E distributions. For F{sub tot}, the standard deviations are 1.2% for HELIOS, 0.9% for CASMO-4, 0.8% for MCNP4B, and 1.7% for BOXER. Standard deviations of 1.1% for HELIOS, CASMO-4, and MCNP4B and 1.2% for BOXER were obtained in the case of C{sub 8}. Despite the high degree of accuracy observed on the average, it was found that the five burnable-absorber fuel pins investigated showed a noticeable underprediction of F{sub tot}, quite systematically, for the deterministic codes evaluated (average C/E for the burnable-absorber fuel pins in the range 0.974 to 0.988, depending on the code)« less

Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mays, Brian; Jackson, R. Brian

2017-03-08

The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services.more » The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.« less

Colossal Tooling Design: 3D Simulation for Ergonomic Analysis

NASA Technical Reports Server (NTRS)

Hunter, Steve L.; Dischinger, Charles; Thomas, Robert E.; Babai, Majid

2003-01-01

The application of high-level 3D simulation software to the design phase of colossal mandrel tooling for composite aerospace fuel tanks was accomplished to discover and resolve safety and human engineering problems. The analyses were conducted to determine safety, ergonomic and human engineering aspects of the disassembly process of the fuel tank composite shell mandrel. Three-dimensional graphics high-level software, incorporating various ergonomic analysis algorithms, was utilized to determine if the process was within safety and health boundaries for the workers carrying out these tasks. In addition, the graphical software was extremely helpful in the identification of material handling equipment and devices for the mandrel tooling assembly/disassembly process.

24 CFR 3285.601 - Field assembly.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 24 Housing and Urban Development 5 2011-04-01 2011-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

24 CFR 3285.601 - Field assembly.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 24 Housing and Urban Development 5 2013-04-01 2013-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

24 CFR 3285.601 - Field assembly.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 24 Housing and Urban Development 5 2014-04-01 2014-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

24 CFR 3285.601 - Field assembly.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 24 Housing and Urban Development 5 2012-04-01 2012-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Tianfu; Ma, Zhuang; Li, Guoping

Electrostatic self-assembly in organic solvent without intensively oxidative or corrosive environments, was adopted to prepare Al/Fe{sub 2}O{sub 3}/MWCNT nanostructured energetic materials as an energy generating material. The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles. This spontaneous assembly method without any surfactant chemistry or other chemical and biological moieties decreased the aggregation of the same nanoparticles largely, moreover, the poor interfacial contact between the Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles was improved significantly, which was the key characteristic ofmore » high performance nanostructured energetic materials. In addition, the assembly process was confirmed as Diffusion-Limited Aggregation. The assembled Al/Fe{sub 2}O{sub 3}/MWCNT nanostructured energetic materials showed excellent performance with heat release of 2400 J/g, peak pressure of 0.42 MPa and pressurization rate of 105.71 MPa/s, superior to that in the control group Al/Fe{sub 2}O{sub 3} nanostructured energetic materials prepared by sonication with heat release of 1326 J/g, peak pressure of 0.19 MPa and pressurization rate of 33.33 MPa/s. Therefore, the approach, which is facile, opens a promising route to the high performance nanostructured energetic materials. - Graphical abstract: The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles. - Highlights: • A facile spontaneous electrostatic assembly strategy without surfactant was adopted. • The fuels and oxidizers assembled into densely packed nanostructured composites. • The assembled nanostructured energetic materials have excellent performance. • This high performance energetic material can be scaled up for practical application. • This strategy can be applied into other nanostructured energetic material systems.« less

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

NASA Astrophysics Data System (ADS)

Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

2018-02-01

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

ADDING REALISM TO NUCLEAR MATERIAL DISSOLVING ANALYSIS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, B.

2011-08-15

Two new criticality modeling approaches have greatly increased the efficiency of dissolver operations in H-Canyon. The first new approach takes credit for the linear, physical distribution of the mass throughout the entire length of the fuel assembly. This distribution of mass is referred to as the linear density. Crediting the linear density of the fuel bundles results in using lower fissile concentrations, which allows higher masses to be charged to the dissolver. Also, this approach takes credit for the fact that only part of the fissile mass is wetted at a time. There are multiple assemblies stacked on top ofmore » each other in a bundle. On average, only 50-75% of the mass (the bottom two or three assemblies) is wetted at a time. This means that only 50-75% (depending on operating level) of the mass is moderated and is contributing to the reactivity of the system. The second new approach takes credit for the progression of the dissolving process. Previously, dissolving analysis looked at a snapshot in time where the same fissile material existed both in the wells and in the bulk solution at the same time. The second new approach models multiple consecutive phases that simulate the fissile material moving from a high concentration in the wells to a low concentration in the bulk solution. This approach is more realistic and allows higher fissile masses to be charged to the dissolver.« less

Premixer assembly for mixing air and fuel for combustion

DOEpatents

York, William David; Johnson, Thomas Edward; Keener, Christopher Paul

2016-12-13

A premixer assembly for mixing air and fuel for combustion includes a plurality of tubes disposed at a head end of a combustor assembly. Also included is a tube of the plurality of tubes, the tube including an inlet end and an outlet end. Further included is at least one non-circular portion of the tube extending along a length of the tube, the at least one non-circular portion having a non-circular cross-section, and the tube having a substantially constant cross-sectional area along its length

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, C. S.; Zhang, Hongbin

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M. M.

1998-10-14

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less

HSPES membrane electrode assembly

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor); Yen, Shiao-Ping (Inventor)

2000-01-01

An improved fuel cell electrode, as well as fuel cells and membrane electrode assemblies that include such an electrode, in which the electrode includes a backing layer having a sintered layer thereon, and a non-sintered free-catalyst layer. The invention also features a method of forming the electrode by sintering a backing material with a catalyst material and then applying a free-catalyst layer.

5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND LOWER FUEL ELEMENTS. TAKEN FROM TOP OF SHIELDING TANK WITH CAMERA POINTING TOWARDS BOTTOM OF TANK. SHOWS LADDER, SQUARE LIFTING FRAME, FUEL ELEMENT HOLDERS, AND CABLE CYLINDERS. INEL PHOTO NUMBER 65-5434, TAKEN OCTOBER 20, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Validation of the analytical methods in the LWR code BOXER for gadolinium-loaded fuel pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paratte, J.M.; Arkuszewski, J.J.; Kamboj, B.K.

1990-01-01

Due to the very high absorption occurring in gadolinium-loaded fuel pins, calculations of lattices with such pins present are a demanding test of the analysis methods in light water reactor (LWR) cell and assembly codes. Considerable effort has, therefore, been devoted to the validation of code methods for gadolinia fuel. The goal of the work reported in this paper is to check the analysis methods in the LWR cell/assembly code BOXER and its associated cross-section processing code ETOBOX, by comparison of BOXER results with those from a very accurate Monte Carlo calculation for a gadolinium benchmark problem. Initial results ofmore » such a comparison have been previously reported. However, the Monte Carlo calculations, done with the MCNP code, were performed at Los Alamos National Laboratory using ENDF/B-V data, while the BOXER calculations were performed at the Paul Scherrer Institute using JEF-1 nuclear data. This difference in the basic nuclear data used for the two calculations, caused by the restricted nature of these evaluated data files, led to associated uncertainties in a comparison of the results for methods validation. In the joint investigations at the Georgia Institute of Technology and PSI, such uncertainty in this comparison was eliminated by using ENDF/B-V data for BOXER calculations at Georgia Tech.« less

Methods and systems to thermally protect fuel nozzles in combustion systems

DOEpatents

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

2013-12-17

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

Green-fuel-mediated synthesis of self-assembled NiO nano-sticks for dual applications—photocatalytic activity on Rose Bengal dye and antimicrobial action on bacterial strains

NASA Astrophysics Data System (ADS)

Iyyappa Rajan, P.; Vijaya, J. Judith; Jesudoss, S. K.; Kaviyarasu, K.; Kennedy, L. John; Jothiramalingam, R.; Al-Lohedan, Hamad A.; Vaali-Mohammed, Mansoor-Ali

2017-08-01

With aim of promoting the employability of green fuels in the synthesis of nano-scaled materials with new kinds of morphologies for multiple applications, successful synthesis of self-assembled NiO nano-sticks was achieved through a 100% green-fuel-mediated hot-plate combustion reaction. The synthesized NiO nano-sticks show excellent photocatalytic activity on Rose Bengal dye and superior antibacterial potential towards both Gram-positive and Gram-negative bacteria.

Advanced membrane electrode assemblies for fuel cells

DOEpatents

Kim, Yu Seung; Pivovar, Bryan S.

2012-07-24

A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

General view of a Space Shuttle Main Engine (SSME) mounted ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

General view of a Space Shuttle Main Engine (SSME) mounted on an SSME engine handler, taken in the SSME Processing Facility at Kennedy Space Center. The most prominent features of the engine assembly in this view are the Low-Pressure Fuel Turbopump Discharge Duct looping diagonally across the top of the assembly and connecting to the High-Pressure Fuel Turbopump, the Low-Pressure Oxidizer Turbopump (LPOTP) located center right of the assembly and the LPOTP Discharge Duct looping around from the pump to the underside of the engine assembly in this view. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

Advanced membrane electrode assemblies for fuel cells

DOEpatents

Kim, Yu Seung; Pivovar, Bryan S

2014-02-25

A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

NASA Astrophysics Data System (ADS)

Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

2015-07-01

During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

Neutronic fuel element fabrication

DOEpatents

Korton, George

2004-02-24

This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

Sustained Recycle in Light Water and Sodium-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven J. Piet; Samuel E. Bays; Michael A. Pope

2010-11-01

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in freshmore » fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.« less

Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system

DOEpatents

Christiansen, David W.; Smith, Bob G.

1982-01-01

A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lafleur, Adrienne M.; Ulrich, Timothy J. II; Menlove, Howard O.

Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio wasmore » more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.« less

Calculated Drag of an Aerial Refueling Assembly Through Airplane Performance Analysis

NASA Technical Reports Server (NTRS)

Vachon, Michael Jacob; Ray, Ronald J.

2004-01-01

The aerodynamic drag of an aerial refueling assembly was calculated during the Automated Aerial Refueling project at the NASA Dryden Flight Research Center. An F/A-18A airplane was specially instrumented to obtain accurate fuel flow measurements and to determine engine thrust. A standard Navy air refueling store with a retractable refueling hose and paradrogue was mounted to the centerline pylon of the F/A-18A airplane. As the paradrogue assembly was deployed and stowed, changes in the calculated thrust of the airplane occurred and were equated to changes in vehicle drag. These drag changes were attributable to the drag of the paradrogue assembly. The drag of the paradrogue assembly was determined to range from 200 to 450 lbf at airspeeds from 170 to 250 KIAS. Analysis of the drag data resulted in a single drag coefficient of 0.0056 for the paradrogue assembly that adequately matched the calculated drag for all flight conditions. The drag relief provided to the tanker airplane when a receiver airplane engaged the paradrogue is also documented from 35 to 270 lbf at the various flight conditions tested. The results support the development of accurate aerodynamic models to be used in refueling simulations and control laws for fully autonomous refueling.

Performance assessment of self-interrogation neutron resonance densitometry for spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Hu, Jianwei; Tobin, Stephen J.; LaFleur, Adrienne M.; Menlove, Howard O.; Swinhoe, Martyn T.

2013-11-01

Self-Interrogation Neutron Resonance Densitometry (SINRD) is one of several nondestructive assay (NDA) techniques being integrated into systems to measure spent fuel as part of the Next Generation Safeguards Initiative (NGSI) Spent Fuel Project. The NGSI Spent Fuel Project is sponsored by the US Department of Energy's National Nuclear Security Administration to measure plutonium in, and detect diversion of fuel pins from, spent nuclear fuel assemblies. SINRD shows promising capability in determining the 239Pu and 235U content in spent fuel. SINRD is a relatively low-cost and lightweight instrument, and it is easy to implement in the field. The technique makes use of the passive neutron source existing in a spent fuel assembly, and it uses ratios between the count rates collected in fission chambers that are covered with different absorbing materials. These ratios are correlated to key attributes of the spent fuel assembly, such as the total mass of 239Pu and 235U. Using count rate ratios instead of absolute count rates makes SINRD less vulnerable to systematic uncertainties. Building upon the previous research, this work focuses on the underlying physics of the SINRD technique: quantifying the individual impacts on the count rate ratios of a few important nuclides using the perturbation method; examining new correlations between count rate ratio and mass quantities based on the results of the perturbation study; quantifying the impacts on the energy windows of the filtering materials that cover the fission chambers by tallying the neutron spectra before and after the neutrons go through the filters; and identifying the most important nuclides that cause cooling-time variations in the count rate ratios. The results of these studies show that 235U content has a major impact on the SINRD signal in addition to the 239Pu content. Plutonium-241 and 241Am are the two main nuclides responsible for the variation in the count rate ratio with cooling time. In short, this work provides insights into some of the main factors that affect the performance of SINRD, and it should help improve the hardware design and the algorithm used to interpret the signal for the SINRD technique. In addition, the modeling and simulation techniques used in this work can be easily adopted for analysis of other NDA systems, especially when complex systems like spent nuclear fuel are involved. These studies were conducted at Los Alamos National Laboratory.

Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

NASA Astrophysics Data System (ADS)

Tobin, S. J.; Menlove, H. O.; Swinhoe, M. T.; Schear, M. A.

2011-10-01

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is "technology driven" in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE PAGES

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

2017-03-27

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Analysis of Adsorbed Natural Gas Tank Technology

NASA Astrophysics Data System (ADS)

Knight, Ernest; Schultz, Conrad; Rash, Tyler; Dohnke, Elmar; Stalla, David; Gillespie, Andrew; Sweany, Mark; Seydel, Florian; Pfeifer, Peter

With gasoline being an ever decreasing finite resource and with the desire to reduce humanity's carbon footprint, there has been an increasing focus on innovation of alternative fuel sources. Natural gas burns cleaner, is more abundant, and conforms to modern engines. However, storing compressed natural gas (CNG) requires large, heavy gas cylinders, which limits space and fuel efficiency. Adsorbed natural gas (ANG) technology allows for much greater fuel storage capacity and the ability to store the gas at a much lower pressure. Thus, ANG tanks are much more flexible in terms of their size, shape, and weight. Our ANG tank employs monolithic nanoporous activated carbon as its adsorbent material. Several different configurations of this Flat Panel Tank Assembly (FPTA) along with a Fuel Extraction System (FES) were examined to compare with the mass flow rate demands of an engine.

Numerical Simulations on the Laser Spot Welding of Zirconium Alloy Endplate for Nuclear Fuel Bundle Assembly

NASA Astrophysics Data System (ADS)

Satyanarayana, G.; Narayana, K. L.; Boggarapu, Nageswara Rao

2018-03-01

In the nuclear industry, a critical welding process is joining of an end plate to a fuel rod to form a fuel bundle. Literature on zirconium welding in such a critical operation is limited. A CFD model is developed and performed for the three-dimensional non-linear thermo-fluid analysis incorporating buoyancy and Marnangoni stress and specifying temperature dependent properties to predict weld geometry and temperature field in and around the melt pool of laser spot during welding of a zirconium alloy E110 endplate with a fuel rod. Using this method, it is possible to estimate the weld pool dimensions for the specified laser power and laser-on-time. The temperature profiles will estimate the HAZ and microstructure. The adequacy of generic nature of the model is validated with existing experimental data.

Life cycle design metrics for energy generation technologies: Method, data, and case study

NASA Astrophysics Data System (ADS)

Cooper, Joyce; Lee, Seung-Jin; Elter, John; Boussu, Jeff; Boman, Sarah

A method to assist in the rapid preparation of Life Cycle Assessments of emerging energy generation technologies is presented and applied to distributed proton exchange membrane fuel cell systems. The method develops life cycle environmental design metrics and allows variations in hardware materials, transportation scenarios, assembly energy use, operating performance and consumables, and fuels and fuel production scenarios to be modeled and comparisons to competing systems to be made. Data and results are based on publicly available U.S. Life Cycle Assessment data sources and are formulated to allow the environmental impact weighting scheme to be specified. A case study evaluates improvements in efficiency and in materials recycling and compares distributed proton exchange membrane fuel cell systems to other distributed generation options. The results reveal the importance of sensitivity analysis and system efficiency in interpreting case studies.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lombardo, N.J.; Marseille, T.J.; White, M.D.

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic inmore » form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.« less

Materials and Fuels Complex Tour

ScienceCinema

Miley, Don

2017-12-11

The Materials and Fuels Complex at Idaho National Laboratory is home to several facilities used for the research and development of nuclear fuels. Stops include the Fuel Conditioning Facility, the Hot Fuel Examination Facility (post-irradiation examination), and the Space and Security Power System Facility, where radioisotope thermoelectric generators (RTGs) are assembled for deep space missions.

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2014 CFR

2014-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2012 CFR

2012-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2011 CFR

2011-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2013 CFR

2013-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2010 CFR

2010-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

Inlet nozzle assembly

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Precechtel, Donald R.; Smith, Bob G.; Knight, Ronald C.

1987-01-01

An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

Inlet nozzle assembly

DOEpatents

Christiansen, D.W.; Karnesky, R.A.; Knight, R.C.; Precechtel, D.R.; Smith, B.G.

1985-09-09

An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

DOE PAGES

Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.; ...

2015-10-09

Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.

Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

NASA Astrophysics Data System (ADS)

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

2017-07-01

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

DOEpatents

Stengel, F.G.

1963-12-24

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

High performance internal reforming unit for high temperature fuel cells

DOEpatents

Ma, Zhiwen [Sandy Hook, CT; Venkataraman, Ramakrishnan [New Milford, CT; Novacco, Lawrence J [Brookfield, CT

2008-10-07

A fuel reformer having an enclosure with first and second opposing surfaces, a sidewall connecting the first and second opposing surfaces and an inlet port and an outlet port in the sidewall. A plate assembly supporting a catalyst and baffles are also disposed in the enclosure. A main baffle extends into the enclosure from a point of the sidewall between the inlet and outlet ports. The main baffle cooperates with the enclosure and the plate assembly to establish a path for the flow of fuel gas through the reformer from the inlet port to the outlet port. At least a first directing baffle extends in the enclosure from one of the sidewall and the main baffle and cooperates with the plate assembly and the enclosure to alter the gas flow path. Desired graded catalyst loading pattern has been defined for optimized thermal management for the internal reforming high temperature fuel cells so as to achieve high cell performance.

Preparation of a self-humidifying membrane electrode assembly for fuel cell and its performance analysis

NASA Astrophysics Data System (ADS)

Wang, Cheng; Mao, Zongqiang; Xu, Jingming; Xie, Xiaofeng; Yang, Lizhai

2003-10-01

A novel nano-porous material SiO2-gel was prepared. After being purified by H2O2, then protonized by H2SO4 and desiccated in vacuum, the SiO2-gel, mixed with Nafion solution, was coated between an electrode and a solid electrolyte, which made a new type of self-humidifying membrane electrode assembly. The SiO2 powder was characterized by FTIR, BET and XRD. The surface of the electrodes was characterized by SEM and EDS. The performances of the self-humidifying membrane electrodes were analyzed by polarization discharge and AC impedance under the operation modes of external humidification and self-humidification respectively. Experimental-results indicated that the SiO2 powder held super-hydrophilicity, and the layer of SiO2 and Nafion polymer between electrode and solid electrolyte expanded three-dimension electrochemistry reac-tion area, maintained stability of catalyst layer and enhanced back-diffusion of water from cathode to anode, so the PEM Fuel cell can generate electricity at self-humidification mode. The power density of single PEM fuel cell reached 1.5 W/cm2 under 0.2 Mpa, 70°C and dry hydrogen and oxygen.

Road user fee task force report to the 72nd Oregon Legislative Assembly on the possible alternatives of the current system of taxing highway use through motor vehicle fuel taxes

DOT National Transportation Integrated Search

2003-03-01

Recognizing that the fuel tax is a declining revenue source for Oregon's road system, the 2001 Oregon Legislative Assembly passed House Bill 3946, mandating the formation of the Road User Fee Task Force with the mission to develop a design for revenu...

Manifold seal for fuel cell stack assembly

DOEpatents

Schmitten, Phillip F.; Wright, Maynard K.

1989-01-01

An assembly for sealing a manifold to a stack of fuel cells includes a first resilient member for providing a first sealing barrier between the manifold and the stack. A second resilient member provides a second sealing barrier between the manifold and the stack. The first and second resilient members are retained in such a manner as to define an area therebetween adapted for retaining a sealing composition.

Constitutive and damage material modeling in a high pressure hydrogen environment

NASA Technical Reports Server (NTRS)

Russell, D. A.; Fritzemeier, L. G.

1991-01-01

Numerous components in reusable space propulsion systems such as the SSME are exposed to high pressure gaseous hydrogen environments. Flow areas and passages in the fuel turbopump, fuel and oxidizer preburners, main combustion chamber, and injector assembly contain high pressure hydrogen either high in purity or as hydrogen rich steam. Accurate constitutive and damage material models applicable to high pressure hydrogen environments are therefore needed for engine design and analysis. Existing constitutive and cyclic crack initiation models were evaluated only for conditions of oxidizing environments. The main objective is to evaluate these models for applicability to high pressure hydrogen environments.

The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eccleston, G.W.; Menlove, H.O.; Abhold, M.

1998-12-31

The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 tomore » 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.« less

Downhole steam generator with improved preheating/cooling features

DOEpatents

Donaldson, A. Burl; Hoke, Donald E.; Mulac, Anthony J.

1983-01-01

An apparatus for downhole steam generation employing dual-stage preheaters for liquid fuel and for the water. A first heat exchange jacket for the fuel surrounds the fuel/oxidant mixing section of the combustor assembly downstream of the fuel nozzle and contacts the top of the combustor unit of the combustor assembly, thereby receiving heat directly from the combustion of the fuel/oxidant. A second stage heat exchange jacket surrounds an upper portion of the oxidant supply line adjacent the fuel nozzle receiving further heat from the compression heat which results from pressurization of the oxidant. The combustor unit includes an inner combustor sleeve whose inner wall defines the combustion zone. The inner combustor sleeve is surrounded by two concentric water channels, one defined by the space between the inner combustor sleeve and an intermediate sleeve, and the second defined by the space between the intermediate sleeve and an outer cylindrical housing. The channels are connected by an annular passage adjacent the top of the combustor assembly and the countercurrent nature of the water flow provides efficient cooling of the inner combustor sleeve. An annular water ejector with a plurality of nozzles is provided to direct water downwardly into the combustor unit at the boundary of the combustion zone and along the lower section of the intermediate sleeve.

Downhole steam generator with improved preheating/cooling features. [Patent application

DOEpatents

Donaldson, A.B.; Hoke, D.E.; Mulac, A.J.

1980-10-10

An apparatus is described for downhole steam generation employing dual-stage preheaters for liquid fuel and for the water. A first heat exchange jacket for the fuel surrounds the fuel/oxidant mixing section of the combustor assembly downstream of the fuel nozzle and contacts the top of the combustor unit of the combustor assembly, thereby receiving heat directly from the combustion of the fuel/oxidant. A second stage heat exchange jacket surrounds an upper portion of the oxidant supply line adjacent the fuel nozzle receiving further heat from the compression heat which results from pressurization of the oxidant. The combustor unit includes an inner combustor sleeve whose inner wall defines the combustion zone. The inner combustor sleeve is surrounded by two concentric water channels, one defined by the space between the inner combustor sleeve and an intermediate sleeve, and the second defined by the space between the intermediate sleeve and an outer cylindrical housing. The channels are connected by an annular passage adjacent the top of the combustor assembly and the countercurrent nature of the water flow provides efficient cooling of the inner combustor sleeve. An annular water ejector with a plurality of nozzles is provided to direct water downwardly into the combustor unit at the boundary of the combustion zone and along the lower section of the intermediate sleeve.

High energy X-ray CT study on the central void formations and the fuel pin deformations of FBR fuel assemblies

NASA Astrophysics Data System (ADS)

Katsuyama, Kozo; Nagamine, Tsuyoshi; Matsumoto, Shin-ichiro; Sato, Seichi

2007-02-01

The central void formations and deformations of fuel pins were investigated in fuel assemblies irradiated to high burn-up, using a non-destructive X-ray CT (computer tomography) technique. In this X-ray CT, the effect of strong gamma ray activity could be reduced to a negligible degree by using the pulse of a high energy X-ray source and detecting the intensity of the transmitted X-rays in synchronization with the generated X-rays. Clear cross-sectional images of fuel assemblies irradiated to high burn-up in a fast breeder reactor were successively obtained, in which the wrapping wires, cladding, pellets and central voids could be distinctly seen. The diameter of a typical central void measured by X-ray CT agreed with the one obtained by ceramography within an error of 0.1 mm. Based on this result, the dependence of the central void diameter on the linear heating rate was analyzed. In addition, the deformation behavior of a fuel pin along its axial direction could be analyzed from 20 stepwise X-ray cross-sectional images obtained in a small interval, and the results obtained showed a good agreement with the predictions calculated by two computer codes.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, M.S.; Zawodzinski, C.

1998-08-25

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field there between for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells. 11 figs.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, Mahlon S.; Zawodzinski, Christine

2001-01-01

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field therebetween for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, Mahlon S.; Zawodzinski, Christine

1998-01-01

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field therebetween for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells.

Dismantlement of the TSF-SNAP Reactor Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

Generator module architecture for a large solid oxide fuel cell power plant

DOEpatents

Gillett, James E.; Zafred, Paolo R.; Riggle, Matthew W.; Litzinger, Kevin P.

2013-06-11

A solid oxide fuel cell module contains a plurality of integral bundle assemblies, the module containing a top portion with an inlet fuel plenum and a bottom portion receiving air inlet feed and containing a base support, the base supports dense, ceramic exhaust manifolds which are below and connect to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the fuel cells comprise a fuel cell stack bundle all surrounded within an outer module enclosure having top power leads to provide electrical output from the stack bundle, where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all 100% of the weight of the stack, and each bundle assembly has its own control for vertical and horizontal thermal expansion control.

Fuel cell assembly unit for promoting fluid service and electrical conductivity

DOEpatents

Jones, Daniel O.

1999-01-01

Fluid service and/or electrical conductivity for a fuel cell assembly is promoted. Open-faced flow channel(s) are formed in a flow field plate face, and extend in the flow field plate face between entry and exit fluid manifolds. A resilient gas diffusion layer is located between the flow field plate face and a membrane electrode assembly, fluidly serviced with the open-faced flow channel(s). The resilient gas diffusion layer is restrained against entering the open-faced flow channel(s) under a compressive force applied to the fuel cell assembly. In particular, a first side of a support member abuts the flow field plate face, and a second side of the support member abuts the resilient gas diffusion layer. The support member is formed with a plurality of openings extending between the first and second sides of the support member. In addition, a clamping pressure is maintained for an interface between the resilient gas diffusion layer and a portion of the membrane electrode assembly. Preferably, the support member is spikeless and/or substantially flat. Further, the support member is formed with an electrical path for conducting current between the resilient gas diffusion layer and position(s) on the flow field plate face.

Method for recovering catalytic elements from fuel cell membrane electrode assemblies

DOEpatents

Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

2012-06-26

A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

Fail-safe storage rack for irradiated fuel rod assemblies

DOEpatents

Lewis, D.R.

1993-03-23

A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

Fail-safe storage rack for irradiated fuel rod assemblies

DOEpatents

Lewis, Donald R.

1993-01-01

A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

Investigations of direct methanol fuel cell (DMFC) fading mechanisms

NASA Astrophysics Data System (ADS)

Sarma, Loka Subramanyam; Chen, Ching-Hsiang; Wang, Guo-Rung; Hsueh, Kan-Lin; Huang, Chiou-Ping; Sheu, Hwo-Shuenn; Liu, Ding-Goa; Lee, Jyh-Fu; Hwang, Bing-Joe

In this report, we present the microscopic investigations on various fading mechanisms of a direct methanol fuel cell (DMFC). High energy X-ray diffraction (XRD), X-ray absorption spectroscopy (XAS), energy dispersive X-ray spectroscopy (EDX), and Raman spectroscopic analysis were applied to a membrane-electrode-assembly (MEA) before and after fuel cell operation to figure out the various factors causing its fading. High energy XRD analysis of the fresh and faded MEA revealed that the agglomeration of the catalyst particles in the cathode layer of the faded MEA was more significant than in the anode layer of the faded MEA. The XAS analysis demonstrated that the alloying extent of Pt (J Pt) and Ru (J Ru) in the anode catalyst was increased and decreased, respectively, from the fresh to the faded MEA, indicating that the Ru environment in the anode catalyst was significantly changed after the fuel cell operation. Based on the X-ray absorption edge jump measurements at the Ru K-edge on the anode catalyst of the fresh and the faded MEA it was found that Ru was dissolved from the Pt-Ru catalyst after the fuel cell operation. Both the Ru K-edge XAS and EDX analysis on the cathode catalyst layer of the faded MEA confirms the presence of Ru environment in the cathode catalyst due to the Ru crossover from the anode to the cathode side. The changes in the membrane and the gas diffusion layer (GDL) after the fuel cell operation were observed from the Raman spectroscopy analysis.

Management of thermal peaking factors in CONFU-B PWR assemblies using neutron poisons and tailored enrichment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Visosky, M.; Hejzlar, P.; Kazimi, M.

2006-07-01

CONFU-B assemblies are PWR assemblies containing standard Uranium fuel rods and TRU bearing inert material fuel rods and are designed to achieve net TRU destruction over a 4.5-year irradiation. These highly heterogeneous assemblies tend to exhibit large intra-assembly power peaking factors (IAPPF). Neutronic strategies to reduce IAPPF are developed. The IAPPF are calculated at the assembly level using CASMO4, and these are used to calculate the most restrictive thermal margin (the Minimum Departure from Nucleate Boiling Ratio, MDNBR) using a whole-core VIPRE-01 model. This paper examines two strategies to manage the thermal margin of a CONFU-B assembly while retaining themore » TRU destruction performance: use of neutron poisons and tailored enrichment schemes. Burnable poisons can be used to suppress BOL reactivity of fresh CONFU-B assemblies with only minor impact on MDNBR and TRU destruction performance. Tailored enrichment, along with the use of soluble boron, can achieve significant improvements in MDNBR, but at some cost to TRU destruction performance. (authors)« less

AHTR Mechanical, Structural, And Neutronic Preconceptual Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Varma, Venugopal Koikal; Holcomb, David Eugene; Peretz, Fred J

2012-10-01

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming a commercial reactor class. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The AHTR employs plate type coated particle fuel assemblies with rapid,more » off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month 2-batch cycle with 9 weight-percent enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The present design intent is for used fuel to be stored inside of containment for at least 6 months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates multiple levels of radioactive material containment including fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents. Key building design elements include: 1) below grade siting to minimize vulnerability to aircraft impact, 2) multiple natural circulation decay heat rejection chimneys, 3) seismic base isolation, and 4) decay heat powered back-up electricity generation. The report provides a preconceptual design of the manipulators, the fuel transfer system, and the salt transfer loops. The mechanical handling of the fuel and how it is accomplished without instrumentation inside the salt is described within the report. All drives for the manipulators reside outside the reactor top flange. The design has also taken into account the transportability of major components and how they will be assembled on site« less

Proton exchange membrane fuel cells for space and electric vehicle applications: From basic research to technology development

NASA Technical Reports Server (NTRS)

Srinivasan, Supramaniam; Mukerjee, Sanjeev; Parthasarathy, A.; CesarFerreira, A.; Wakizoe, Masanobu; Rho, Yong Woo; Kim, Junbom; Mosdale, Renaut A.; Paetzold, Ronald F.; Lee, James

1994-01-01

The proton exchange membrane fuel cell (PEMFC) is one of the most promising electrochemical power sources for space and electric vehicle applications. The wide spectrum of R&D activities on PEMFC's, carried out in our Center from 1988 to date, is as follows (1) Electrode Kinetic and Electrocatalysis of Oxygen Reduction; (2) Optimization of Structures of Electrodes and of Membrane and Electrode Assemblies; (3) Selection and Evaluation of Advanced Proton Conducting Membranes and of Operating Conditions to Attain High Energy Efficiency; (4) Modeling Analysis of Fuel Cell Performance and of Thermal and Water Management; and (5) Engineering Design and Development of Multicell Stacks. The accomplishments on these tasks may be summarized as follows: (1) A microelectrode technique was developed to determine the electrode kinetic parameters for the fuel cell reactions and mass transport parameters for the H2 and O2 reactants in the proton conducting membrane. (2) High energy efficiencies and high power densities were demonstrated in PEMFCs with low platinum loading electrodes (0.4 mg/cm(exp 2) or less), advanced membranes and optimized structures of membrane and electrode assemblies, as well as operating conditions. (3) The modeling analyses revealed methods to minimize mass transport limitations, particularly with air as the cathodic reactant; and for efficient thermal and water management. (4) Work is in progress to develop multi-kilowatt stacks with the electrodes containing low platinum loadings.

Nuclear fuel element nut retainer cup. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walton, L.A.

1977-07-19

A typical embodiment has an end fitting for a nuclear reactor fuel element that is joined to the control rod guide tubes by means of a nut plate assembly. The nut plate assembly has an array of nuts, each engaging the respective threaded end of the control rod guide tubes. The nuts, moreover, are retained on the plate during handling and before fuel element assembly by means of hollow cylindrical locking cups that are brazed to the plate and loosely circumscribe the individual enclosed nuts. After the nuts are threaded onto the respective guide tube ends, the locking cups aremore » partially deformed to prevent one or more of the nuts from working loose during reactor operation. The locking cups also prevent loose or broken end fitting parts from becoming entrained in the reactor coolant.« less

Systems and methods for preventing flashback in a combustor assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, Thomas Edward; Ziminsky, Willy Steve; Stevenson, Christian Xavier

2016-04-05

Embodiments of the present application include a combustor assembly. The combustor assembly may include a combustion chamber, a first plenum, a second plenum, and one or more elongate air/fuel premixing injection tubes. Each of the elongate air/fuel premixing injection tubes may include a first length at least partially disposed within the first plenum and configured to receive a first fluid from the first plenum. Moreover, each of the elongate air/fuel premixing injection tubes may include a second length disposed downstream of the first length and at least partially disposed within the second plenum. The second length may be formed ofmore » a porous wall configured to allow a second fluid from the second plenum to enter the second length and create a boundary layer about the porous wall.« less

Membrane electrode gasket assembly (MEGA) technology for polymer electrolyte fuel cells

NASA Astrophysics Data System (ADS)

Pozio, A.; Giorgi, L.; De Francesco, M.; Silva, R. F.; Lo Presti, R.; Danzi, A.

A new technology for the production of a membrane electrode gasket assembly (MEGA) for polymer electrolyte fuel cells (PEFCs) is defined. The MEGA system was prepared by sealing a previously prepared membrane electrode assembly (MEA) in a moulded gasket. For this aim, a proprietary silicone based liquid mixture was injected directly into the MEA borders. Gaskets obtained in different shapes and hardness grades are stable in a wide temperature range. The MEGA technology shows several advantages with respect to traditional PEFCs stack assembling systems: effective membrane saving, reduced fabrication time, possibility of quality control and failed elements substitution. This technology was successfully tested at the ENEA laboratories and the results were acquired in laboratory scale, but industrial production appears to be simple and cheap.

Integral Fast Reactor fuel pin processor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Levinskas, D.

1993-01-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

Integral Fast Reactor fuel pin processor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Levinskas, D.

1993-03-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

In situ observations of water production and distribution in an operating H2/O2 PEM fuel cell assembly using 1H NMR microscopy.

PubMed

Feindel, Kirk W; LaRocque, Logan P-A; Starke, Dieter; Bergens, Steven H; Wasylishen, Roderick E

2004-09-22

Proton NMR imaging was used to investigate in situ the distribution of water in a polymer electrolyte membrane fuel cell operating on H2 and O2. In a single experiment, water was monitored in the gas flow channels, the membrane electrode assembly, and in the membrane surrounding the catalysts. Radial gradient diffusion removes water from the catalysts into the surrounding membrane. This research demonstrates the strength of 1H NMR microscopy as an aid for designing fuel cells to optimize water management.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanan, N. A.; Matos, J. E.

At The request of the Czech Technical University in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. For core configurations C1 and C2, criticality calculations were done for cases with all control rodsmore » at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were done for the C1 core configuration. Finally the reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations.« less

Dry compliant seal for phosphoric acid fuel cell

DOEpatents

Granata, Jr., Samuel J.; Woodle, Boyd M.

1990-01-01

A dry compliant overlapping seal for a phosphoric acid fuel cell preformed f non-compliant Teflon to make an anode seal frame that encircles an anode assembly, a cathode seal frame that encircles a cathode assembly and a compliant seal frame made of expanded Teflon, generally encircling a matrix assembly. Each frame has a thickness selected to accommodate various tolerances of the fuel cell elements and are either bonded to one of the other frames or to a bipolar or end plate. One of the non-compliant frames is wider than the other frames forming an overlap of the matrix over the wider seal frame, which cooperates with electrolyte permeating the matrix to form a wet seal within the fuel cell that prevents process gases from intermixing at the periphery of the fuel cell and a dry seal surrounding the cell to keep electrolyte from the periphery thereof. The frames may be made in one piece, in L-shaped portions or in strips and have an outer perimeter which registers with the outer perimeter of bipolar or end plates to form surfaces upon which flanges of pan shaped, gas manifolds can be sealed.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Ross, Steven; Grey, Carissa Ann

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

IMPROVED TYPE OF FUEL ELEMENT

DOEpatents

Monson, H.O.

1961-01-24

A radiator-type fuel block assembly is described. It has a hexagonal body of neutron fissionable material having a plurality of longitudinal equal- spaced coolant channels therein aligned in rows parallel to each face of the hexagonal body. Each of these coolant channels is hexagonally shaped with the corners rounded and enlarged and the assembly has a maximum temperature isothermal line around each channel which is approximately straight and equidistant between adjacent channels.

Posttest examination of Sodium Loop Safety Facility experiments. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, J.W.

In-reactor, safety experiments performed in the Sodium Loop Safety Facility (SLSF) rely on comprehensive posttest examinations (PTE) to characterize the postirradiation condition of the cladding, fuel, and other test-subassembly components. PTE information and on-line instrumentation data, are analyzed to identify the sequence of events and the severity of the accident for each experiment. Following in-reactor experimentation, the SLSF loop and test assembly are transported to the Hot Fuel Examination Facility (HFEF) for initial disassembly. Goals of the HFEF-phase of the PTE are to retrieve the fuel bundle by dismantling the loop and withdrawing the test assembly, to assess the macro-conditionmore » of the fuel bundle by nondestructive examination techniques, and to prepare the fuel bundle for shipment to the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory.« less

Horizontal modular dry irradiated fuel storage system

DOEpatents

Fischer, Larry E.; McInnes, Ian D.; Massey, John V.

1988-01-01

A horizontal, modular, dry, irradiated fuel storage system (10) includes a thin-walled canister (12) for containing irradiated fuel assemblies (20), which canister (12) can be positioned in a transfer cask (14) and transported in a horizontal manner from a fuel storage pool (18), to an intermediate-term storage facility. The storage system (10) includes a plurality of dry storage modules (26) which accept the canister (12) from the transfer cask (14) and provide for appropriate shielding about the canister (12). Each module (26) also provides for air cooling of the canister (12) to remove the decay heat of the irradiated fuel assemblies (20). The modules (26) can be interlocked so that each module (26) gains additional shielding from the next adjacent module (26). Hydraulic rams (30) are provided for inserting and removing the canisters (12) from the modules (26).

Durability test with fuel starvation using a Pt/CNF catalyst in PEMFC.

PubMed

Jung, Juhae; Park, Byungil; Kim, Junbom

2012-01-05

In this study, a catalyst was synthesized on carbon nanofibers [CNFs] with a herringbone-type morphology. The Pt/CNF catalyst exhibited low hydrophilicity, low surface area, high dispersion, and high graphitic behavior on physical analysis. Electrodes (5 cm2) were prepared by a spray method, and the durability of the Pt/CNF was evaluated by fuel starvation. The performance was compared with a commercial catalyst before and after accelerated tests. The fuel starvation caused carbon corrosion with a reverse voltage drop. The polarization curve, EIS, and cyclic voltammetry were analyzed in order to characterize the electrochemical properties of the Pt/CNF. The performance of a membrane electrode assembly fabricated from the Pt/CNF was maintained, and the electrochemical surface area and cell resistance showed the same trend. Therefore, CNFs are expected to be a good support in polymer electrolyte membrane fuel cells.

Durability test with fuel starvation using a Pt/CNF catalyst in PEMFC

PubMed Central

2012-01-01

In this study, a catalyst was synthesized on carbon nanofibers [CNFs] with a herringbone-type morphology. The Pt/CNF catalyst exhibited low hydrophilicity, low surface area, high dispersion, and high graphitic behavior on physical analysis. Electrodes (5 cm2) were prepared by a spray method, and the durability of the Pt/CNF was evaluated by fuel starvation. The performance was compared with a commercial catalyst before and after accelerated tests. The fuel starvation caused carbon corrosion with a reverse voltage drop. The polarization curve, EIS, and cyclic voltammetry were analyzed in order to characterize the electrochemical properties of the Pt/CNF. The performance of a membrane electrode assembly fabricated from the Pt/CNF was maintained, and the electrochemical surface area and cell resistance showed the same trend. Therefore, CNFs are expected to be a good support in polymer electrolyte membrane fuel cells. PMID:22221426

Low Pressure Nuclear Thermal Rocket (LPNTR) concept

NASA Technical Reports Server (NTRS)

Ramsthaler, J. H.

1991-01-01

A background and a description of the low pressure nuclear thermal system are presented. Performance, mission analysis, development, critical issues, and some conclusions are discussed. The following subject areas are covered: LPNTR's inherent advantages in critical NTR requirement; reactor trade studies; reference LPNTR; internal configuration and flow of preliminary LPNTR; particle bed fuel assembly; preliminary LPNTR neutronic study results; multiple LPNTR engine concept; tank and engine configuration for mission analysis; LPNTR reliability potential; LPNTR development program; and LPNTR program costs.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Since much of the fuel is very old, Cerenkov viewing devices are often not very useful even for the top layer. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times canmore » exist in any given pool. A spent fuel neutron counting tool consisting of a fission chamber, SFNC, has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups to levels up 11,000 MWd/t, the existing signal processing software of the tool was found to fail due to non-linearity of the source term with burnup. A new Graphical User Interface software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times and establish maps of expected signals at various pool locations. The algorithm employed in the software uses a set of transfer functions in a 47-energy group structure which are coupled with a 47-energy group neutron source spectrum based on various cooling times and burnups for each of the two enrichment levels. The database of the software consists of these transfer functions for the three different inter-assembly pitches that the fuel is stored in at the site. The transfer functions were developed for a 6 by 6 matrix of fuel assemblies with the detector placed at the center surrounded by four near neighbors, eight next nearest neighbors and so on for the 36 assemblies. These calculations were performed using Monte Carlo radiation transport methods. The basic methodology consisted of starting sources in each of the assemblies and tallying the contribution to the detector by a single neutron in each of the 47 energy groups used. Thus for the single existing symmetric pitch in the pools, where the vertical and horizontal separations are equal, only 6 sets of transfer functions are required. For the two asymmetrical pitches, nine sets of transfer functions are stored. In addition, source spectra at burnups ranging from 4000 to 20000 MWd/t and cooling times up to 40 years are stored. These source terms were established based on CANDU 37-rod fuel that is very similar to the Atucha fuel. Linear interpolation is used by the software for both burnup and cooling time to establish source terms at any intermediate condition. Using the burnup, cooling time and initial enrichment of the surrounding assemblies a set of source strengths in the 47-group structure for each of the 36 assemblies is established and multiplied group-wise with the appropriate transfer function set. The grand total over the 47 groups for all 36 assemblies is the predicted signal at the detector. The software was initially calibrated against a set of typically 5-6 measurements chosen from among the measured data at each level of the six pools and calibration factors were established. The set used for calibration is chosen such that it is fairly representative of the range of spent fuel assembly characteristics present in each level. Once established, these calibration factors can be repeatedly used for verification purposes. Recalibration will be required if the hardware or pool configurations has changed. It will also be required if a long enough time has elapsed since they were established thus making a cooling time correction necessary. The objective of the inspection is to detect missing fuel from one or more nearest neighbors of the detector. During the verification mode of the software, the predicted and measured signals are compared and the inspector is alerted if the difference between the two signals is beyond a set tolerance limit. Based on the uncertainties associated with both the calculations and measurements, a lower limit of the tolerance will be 15% with an upper limit of 20%. For the most part a 20% tolerance limit will be able to detect a missing assembly since in the vast majority of cases the drop in signal due to a single missing nearest neighbor assembly will be in the range 24-27%. The software was benchmarked against an extensive set of measured data taken at the site in 2004. Overall, 326 data points were examined and the prediction of the calibrated software was compared to the measurements within a set tolerance of ±20%. Of these, 283 of the predicted signals representing 87% of the total matched the measured data within ±10%. A further 27 or 8% were in the range of ±10-15% and 8 or 2.5% were in the range of ±15-20%. Thus, 97.5% of the data matched the measurements within the set tolerance limit of 20%, with 95% matching measured data with the lowest allowed tolerance limit of ±15%. The remaining 2.5% had measured signals that were very different from those at locations with very similar surrounding assemblies and the cause of these discrepancies could not be ascertained from the measurement logs. In summary, 97.5% of the predictions matched the measurements within the set 20% tolerance limit providing proof of the robustness of the software. This software package linked to SFNC will be deployed at the site and will enhance the capability of gross defect verification for the whole range of burnup, cooling time and initial enrichments of the spent fuel being discharged into the various pools at the Atucha-I reactor site.« less

New approaches for MOX multi-recycling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gain, T.; Bouvier, E.; Grosman, R.

Due to its low fissile content after irradiation, Pu from used MOX fuel is considered by some as not recyclable in LWR (Light Water Reactors). The point of this paper is hence to go back to those statements and provide a new analysis based on AREVA extended experience in the fields of fissile and fertile material management and optimized waste management. This is done using the current US fuel inventory as a case study. MOX Multi-recycling in LWRs is a closed cycle scenario where U and Pu management through reprocessing and recycling leads to a significant reduction of the usedmore » assemblies to be stored. The recycling of Pu in MOX fuel is moreover a way to maintain the self-protection of the Pu-bearing assemblies. With this scenario, Pu content is also reduced repetitively via a multi-recycling of MOX in LWRs. Simultaneously, {sup 238}Pu content decreases. All along this scenario, HLW (High-Level Radioactive Waste) vitrified canisters are produced and planned for deep geological disposal. Contrary to used fuel, HLW vitrified canisters do not contain proliferation materials. Moreover, the reprocessing of used fuel limits the space needed on current interim storage. With MOX multi-recycling in LWR, Pu isotopy needs to be managed carefully all along the scenario. The early introduction of a limited number of SFRs (Sodium Fast Reactors) can therefore be a real asset for the overall system. A few SFRs would be enough to improve the Pu isotopy from used LWR MOX fuel and provide a Pu-isotopy that could be mixed back with multi-recycled Pu from LWRs, hence increasing the Pu multi-recycling potential in LWRs.« less

Manufacturing of Low Cost, Durable Membrane Electrode Assemblies Engineered for Rapid Conditioning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Busby, Colin

Over the past 20 years significant progress in membrane-electrode assembly (MEA) technology development for polymer electrolyte fuel cells (PEMFCs) has resulted in the PEMFC technology approaching a commercial reality for transportation applications. However, there remain two primary technical challenges to be addressed in the MEA. First and foremost is meeting the automotive cost targets: Producing a fuel cell stack cost competitive with today’s internal combustion engine. In addition to the material cost, MEA (and other components) and stack assembly production methods must be amenable for use in low cost, high speed, automotive assembly line. One impediment to this latter goalmore » is that stack components must currently go through a long and tedious conditioning procedure before they produce optimal power. This so-called “break-in” can take many hours, and can involve quite complex voltage, temperature and/or pressure steps. These break-in procedures must be simplified and the time required reduced if fuel cells are to become a viable automotive engine. The second challenge is to achieve the durability targets in real-world automotive duty cycle operations. Significant improvements in cost, break-in time, and durability for the key component of fuel cell stacks, MEAs were achieved in this project. Advanced modeling was used to guide design of the new MEA to maximize performance and durability. A new, innovative process and manufacturing approach utilizing direct in-line coating using scalable, cost-competitive, continuous high volume 3-layer rolled-good manufacturing processes was developed and validated by single cell and short stack testing. In addition, the direct coating methods employed were shown to reduce the cost for sacrificial films. Furthermore, Gore has demonstrated a 10 µm reinforced membrane that is used in the new low-cost process and can meet automotive power density and durability targets. Across a wide range of operating conditions, the direct-coated MEA outperformed the commercial baseline MEA, and did so through a process that delivers MEAs at $92.35/m2 at a volume of 500,000 systems per year, according to Strategic Analysis (SA) estimates.« less

FUEL ROD ASSEMBLY

DOEpatents

Hutter, E.

1959-09-01

A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

77 FR 2658 - Airworthiness Directives; Bombardier, Inc. Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-19

... high pressure fuel lines due to improper installation of an expandable pin on the lower cowl assembly... chafing of the high pressure fuel lines, which if not corrected, could cause fuel leakage in a fire zone... on a high pressure (HP) fuel line. The source of chafing was related to the improper installation of...

Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

DOEpatents

Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

2012-04-24

A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

Nuclear reactors built, being built, or planned, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, B.

1992-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Improving Energy Efficiency for the Vehicle Assembly Industry: A Discrete Event Simulation Approach

NASA Astrophysics Data System (ADS)

Oumer, Abduaziz; Mekbib Atnaw, Samson; Kie Cheng, Jack; Singh, Lakveer

2016-11-01

This paper presented a Discrete Event Simulation (DES) model for investigating and improving energy efficiency in vehicle assembly line. The car manufacturing industry is one of the highest energy consuming industries. Using Rockwell Arena DES package; a detailed model was constructed for an actual vehicle assembly plant. The sources of energy considered in this research are electricity and fuel; which are the two main types of energy sources used in a typical vehicle assembly plant. The model depicts the performance measurement for process- specific energy measures of painting, welding, and assembling processes. Sound energy efficiency model within this industry has two-fold advantage: reducing CO2 emission and cost reduction associated with fuel and electricity consumption. The paper starts with an overview of challenges in energy consumption within the facilities of automotive assembly line and highlights the parameters for energy efficiency. The results of the simulation model indicated improvements for energy saving objectives and reduced costs.

Dimensionless numbers and correlating equations for the analysis of the membrane-gas diffusion electrode assembly in polymer electrolyte fuel cells

NASA Astrophysics Data System (ADS)

Gyenge, E. L.

The Quraishi-Fahidy method [Can. J. Chem. Eng. 59 (1981) 563] was employed to derive characteristic dimensionless numbers for the membrane-electrolyte, cathode catalyst layer and gas diffuser, respectively, based on the model presented by Bernardi and Verbrugge for polymer electrolyte fuel cells [AIChE J. 37 (1991) 1151]. Monomial correlations among dimensionless numbers were developed and tested against experimental and mathematical modeling results. Dimensionless numbers comparing the bulk and surface-convective ionic conductivities, the electric and viscous forces and the current density and the fixed surface charges, were employed to describe the membrane ohmic drop and its non-linear dependence on current density due to membrane dehydration. The analysis of the catalyst layer yielded electrode kinetic equivalents of the second Damköhler number and Thiele modulus, influencing the penetration depth of the oxygen reduction front based on the pseudohomogeneous film model. The correlating equations for the catalyst layer could describe in a general analytical form, all the possible electrode polarization scenarios such as electrode kinetic control coupled or not with ionic and/or oxygen mass transport limitation. For the gas diffusion-backing layer correlations are presented in terms of the Nusselt number for mass transfer in electrochemical systems. The dimensionless number-based correlating equations for the membrane electrode assembly (MEA) could provide a practical approach to quantify single-cell polarization results obtained under a variety of experimental conditions and to implement them in models of the fuel cell stack.

Combustor assembly in a gas turbine engine

DOEpatents

Wiebe, David J; Fox, Timothy A

2013-02-19

A combustor assembly in a gas turbine engine. The combustor assembly includes a combustor device coupled to a main engine casing, a first fuel injection system, a transition duct, and an intermediate duct. The combustor device includes a flow sleeve for receiving pressurized air and a liner disposed radially inwardly from the flow sleeve. The first fuel injection system provides fuel that is ignited with the pressurized air creating first working gases. The intermediate duct is disposed between the liner and the transition duct and defines a path for the first working gases to flow from the liner to the transition duct. An intermediate duct inlet portion is associated with a liner outlet and allows movement between the intermediate duct and the liner. An intermediate duct outlet portion is associated with a transition duct inlet section and allows movement between the intermediate duct and the transition duct.

Isochoric Implosions for Fast Ignition

NASA Astrophysics Data System (ADS)

Clark, Daniel; Tabak, Max

2006-10-01

Various gain models have shown the potentially great advantages of Fast Ignition (FI) Inertial Confinement Fusion (ICF) over its conventional hotspot ignition counterpart. These gain models, however, all assume nearly uniform-density fuel assemblies. By contrast, typical ICF implosions yield hollowed fuel assemblies with a high-density shell of fuel surrounding a low-density, high-pressure hotspot. To realize fully the advantages of FI, then, an alternative implosion design must be found which yields nearly isochoric fuel assemblies without substantial hotspots. Here, it is shown that a self-similar spherical implosion of the type originally studied by Guderley [Luftfahrtforschung 19, 302 (1942)] may be employed to yield precisely such quasi-isochoric imploded states. The difficulty remains, however, of accessing these self-similarly imploding configurations from initial conditions representing an actual ICF target, namely a uniform, solid-density shell at rest. Furthermore, these specialized implosions must be realized for practicable drive parameters, i.e., accessible peak pressures, shell aspect ratios, etc. An implosion scheme is presented which meets all of these requirements, suggesting the possibility of genuinely isochoric implosions for FI.

Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

Single-wall carbon nanotube-based proton exchange membrane assembly for hydrogen fuel cells.

PubMed

Girishkumar, G; Rettker, Matthew; Underhile, Robert; Binz, David; Vinodgopal, K; McGinn, Paul; Kamat, Prashant

2005-08-30

A membrane electrode assembly (MEA) for hydrogen fuel cells has been fabricated using single-walled carbon nanotubes (SWCNTs) support and platinum catalyst. Films of SWCNTs and commercial platinum (Pt) black were sequentially cast on a carbon fiber electrode (CFE) using a simple electrophoretic deposition procedure. Scanning electron microscopy and Raman spectroscopy showed that the nanotubes and the platinum retained their nanostructure morphology on the carbon fiber surface. Electrochemical impedance spectroscopy (EIS) revealed that the carbon nanotube-based electrodes exhibited an order of magnitude lower charge-transfer reaction resistance (R(ct)) for the hydrogen evolution reaction (HER) than did the commercial carbon black (CB)-based electrodes. The proton exchange membrane (PEM) assembly fabricated using the CFE/SWCNT/Pt electrodes was evaluated using a fuel cell testing unit operating with H(2) and O(2) as input fuels at 25 and 60 degrees C. The maximum power density obtained using CFE/SWCNT/Pt electrodes as both the anode and the cathode was approximately 20% better than that using the CFE/CB/Pt electrodes.

Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Chang, G. S.; Lillo, M. A.

2009-08-01

The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y-Z mini-plate fuel model was developed. The Y-Z model divides each fuel plate into 30 equal intervals in both the Y and Z directions. The MCNP-calculated results and the detailed Y-Z fission power mapping were used to help design the AFIP fuel test assembly to demonstrate that the AFIP test assembly thermal-hydraulic limits will not exceed the ATR safety limits.

78 FR 27951 - Foreign-Trade Zone (FTZ) 75-Phoenix, Arizona; Notification of Proposed Production Activity...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-13

... systems; duct temperature limiters; air/oil heat exchangers; oil cooler fans; fuel filter assemblies... assemblies; filter extractors; de- coupler/disassembly wrenches; torque wrench adaptors; test benches; drills...; filter assemblies; oil filter install kits; cartridge screens; filter housings; trim balance weights...

Fuel nozzle assembly for use in turbine engines and methods of assembling same

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2015-02-03

A fuel nozzle for use with a turbine engine is described herein. The fuel nozzle includes a housing that is coupled to a combustor liner defining a combustion chamber. The housing includes an endwall that at least partially defines the combustion chamber. A plurality of mixing tubes extends through the housing for channeling fuel to the combustion chamber. Each mixing tube of the plurality of mixing tubes includes an inner surface that extends between an inlet portion and an outlet portion. The outlet portion is oriented adjacent the housing endwall. At least one of the plurality of mixing tubes includes a plurality of projections that extend outwardly from the outlet portion. Adjacent projections are spaced a circumferential distance apart such that a groove is defined between each pair of circumferentially-apart projections to facilitate enhanced mixing of fuel in the combustion chamber.

Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

DOE Office of Scientific and Technical Information (OSTI.GOV)

F. Delage; J. Carmack; C. B. Lee

2013-10-01

The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxidemore » and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.« less

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, Charles W.

1992-01-01

A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

Spacer grid assembly and locking mechanism

DOEpatents

Snyder, Jr., Harold J.; Veca, Anthony R.; Donck, Harry A.

1982-01-01

A spacer grid assembly is disclosed for retaining a plurality of fuel rods in substantially parallel spaced relation, the spacer grids being formed with rhombic openings defining contact means for engaging from one to four fuel rods arranged in each opening, the spacer grids being of symmetric configuration with their rhombic openings being asymmetrically offset to permit inversion and relative rotation of the similar spacer grids for improved support of the fuel rods. An improved locking mechanism includes tie bars having chordal surfaces to facilitate their installation in slotted circular openings of the spacer grids, the tie rods being rotatable into locking engagement with the slotted openings.

10 CFR 32.2 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Definitions. 32.2 Section 32.2 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL... disposal, or nuclear material contained in any fuel assembly, subassembly, fuel rod, or fuel pellet...

10 CFR 32.2 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Definitions. 32.2 Section 32.2 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL... disposal, or nuclear material contained in any fuel assembly, subassembly, fuel rod, or fuel pellet...

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Gurinsky, D.H.; Powell, R.W.; Fox, M.

1959-11-24

A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

Self-Assembled Nano-energetic Gas Generators based on Bi2O3

NASA Astrophysics Data System (ADS)

Hobosyan, Mkhitar; Trevino, Tyler; Martirosyan, Karen

2012-10-01

Nanoenergetic Gas-Generators are formulations that rapidly release a large amount of gaseous products and generate a fast moving thermal wave. They are mainly based on thermite systems, which are pyrotechnic mixtures of metal powders (fuel- Al, Mg, etc.) and metal oxides (oxidizer, Bi2O3, Fe2O3, WO3, MoO3 etc.) that can generate an exothermic oxidation-reduction reaction referred to as a thermite reaction. A thermite reaction releases a large amount of energy and can generate rapidly extremely high temperatures. The intimate contact between the fuel and oxidizer can be enhanced by use of nano instead of micro particles. The contact area between oxidizer and metal particles depends from method of mixture preparation. In this work we utilize the self-assembly processes, which use the electrostatic forces to produce ordered and self-organized binary systems. In this process the intimate contact significantly enhances and gives the ability to build an energetic material in molecular level, which is crucial for thepressure discharge efficiency of nano-thermites. The DTA-TGA, Zeta-size analysis and FTIR technique were performed to characterize the Bi2O3 particles. The self-assembly of Aluminum and Bi2O3 was conducted in sonic bath with appropriate solvents and linkers. The resultant thermite pressure discharge values were tested in modified Parr reactor. In general, the self-assembled thermites give much higher-pressure discharge values than the thermites prepared with conventional roll-mixing technique.

Upper Stage Flight Experiment 10K Engine Design and Test Results

NASA Technical Reports Server (NTRS)

Ross, R.; Morgan, D.; Crockett, D.; Martinez, L.; Anderson, W.; McNeal, C.

2000-01-01

A 10,000 lbf thrust chamber was developed for the Upper Stage Flight Experiment (USFE). This thrust chamber uses hydrogen peroxide/JP-8 oxidizer/fuel combination. The thrust chamber comprises an oxidizer dome and manifold, catalyst bed assembly, fuel injector, and chamber/nozzle assembly. Testing of the engine was done at NASA's Stennis Space Center (SSC) to verify its performance and life for future upper stage or Reusable Launch Vehicle applications. Various combinations of silver screen catalyst beds, fuel injectors, and combustion chambers were tested. Results of the tests showed high C* efficiencies (97% - 100%) and vacuum specific impulses of 275 - 298 seconds. With fuel film cooling, heating rates were low enough that the silica/quartz phenolic throat experienced minimal erosion. Mission derived requirements were met, along with a perfect safety record.

Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

2018-01-01

Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been necessary to perform a careful design study of the probe geometry. For this, finite element analysis (FEA) has been performed in combination with practical validation tests on representative fuel dummies with machined flaws to find the probe geometry that best detects a hidden flaw. Tests performed thus far show that gaps down to 25 μm thickness can be detected with good repeatability and good discrimination from lift-off signals.

Criticality Safety Evaluation Report CSER-96-019 for Spent Nuclear Fuel (SNF) Processing and Storage Facilities Multi Canister Overpack (MCO)

DOE Office of Scientific and Technical Information (OSTI.GOV)

KESSLER, S.F.

This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operation at the Cold Vacuum Drying Facility,a nd storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the K{sub eff} = 0.95 criticality safety limit. This revision incorporates the analyses for the sampling/weldmore » station in the Canister Storage Building and additional analysis of the MCO during the draining at CVDF. Additional discussion of the scrap basket model was added to show why the addition of copper divider plates was not included in the models.« less

An analysis of the sliding pressure start-up of SCWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, F.; Yang, J.; Li, H.

In this paper, the preliminary sliding pressure start-up system and scheme of supercritical water-cooled reactor in CGNPC (CGN-SCWR) were proposed. Thermal-hydraulic behavior in start-up procedures was analyzed in detail by employing advanced reactor subchannel analysis software ATHAS. The maximum cladding temperature (MCT for short) and core power of fuel assembly during the whole start-up process were investigated comparatively. The results show that the recommended start-up scheme meets the design requirements from the perspective of thermal-hydraulic. (authors)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hannan, N. A.; Matos, J. E.; Stillman, J. A.

At the request of the Czech Technical University (CTU) in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. Fore core configurations C1 and C2, criticality calculations were done for cases with all controlmore » rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were doe for the C1 core configuration. The reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations. Finally, the reactivity feedback coefficients, the prompt neutron lifetime, and the total effective delay neutron fraction were calculated for each of the three cores.« less

Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruggles, A.E.

1990-10-12

The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results aremore » related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.« less

PEM/SPE fuel cell

DOEpatents

Grot, Stephen Andreas

1998-01-01

A PEM/SPE fuel cell including a membrane-electrode assembly (MEA) having a plurality of oriented filament embedded the face thereof for supporting the MEA and conducting current therefrom to contiguous electrode plates.

Fission matrix-based Monte Carlo criticality analysis of fuel storage pools

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farlotti, M.; Ecole Polytechnique, Palaiseau, F 91128; Larsen, E. W.

2013-07-01

Standard Monte Carlo transport procedures experience difficulties in solving criticality problems in fuel storage pools. Because of the strong neutron absorption between fuel assemblies, source convergence can be very slow, leading to incorrect estimates of the eigenvalue and the eigenfunction. This study examines an alternative fission matrix-based Monte Carlo transport method that takes advantage of the geometry of a storage pool to overcome this difficulty. The method uses Monte Carlo transport to build (essentially) a fission matrix, which is then used to calculate the criticality and the critical flux. This method was tested using a test code on a simplemore » problem containing 8 assemblies in a square pool. The standard Monte Carlo method gave the expected eigenfunction in 5 cases out of 10, while the fission matrix method gave the expected eigenfunction in all 10 cases. In addition, the fission matrix method provides an estimate of the error in the eigenvalue and the eigenfunction, and it allows the user to control this error by running an adequate number of cycles. Because of these advantages, the fission matrix method yields a higher confidence in the results than standard Monte Carlo. We also discuss potential improvements of the method, including the potential for variance reduction techniques. (authors)« less

Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

NASA Astrophysics Data System (ADS)

Rochman, Dimitri A.; Vasiliev, Alexander; Dokhane, Abdelhamid; Ferroukhi, Hakim

2018-05-01

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

Method of improving fuel cell performance by removing at least one metal oxide contaminant from a fuel cell electrode

DOEpatents

Kim, Yu Seung [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Zelenay, Piotr [Los Alamos, NM

2009-08-18

A method of removing contaminants from a fuel cell catalyst electrode. The method includes providing a getter electrode and a fuel cell catalyst electrode having at least one contaminant to a bath and applying a voltage sufficient to drive the contaminant from the fuel cell catalyst electrode to the getter electrode. Methods of removing contaminants from a membrane electrode assembly of a fuel cell and of improving performance of a fuel cell are also provided.

Analysis and Implementation of Accident Tolerant Nuclear Fuels

NASA Astrophysics Data System (ADS)

Prewitt, Benjamin Joseph

To improve the reliability and robustness of LWR, accident tolerant nuclear fuels and cladding materials are being developed to possibly replace the current UO2/zirconium system. This research highlights UN and U3Si 2, two of the most favorable accident tolerant fuels being developed. To evaluate the commercial feasiblilty of these fuels, two areas of research were conducted. Chemical fabrication routes for both fuels were investigated in detail, considering UO2 and UF6 as potential starting materials. Potential pathways for industrial scale fabrication using these methods were discussed. Neutronic performance of 70%UN-30%U3Si2 composite was evaluated in MNCP using PWR assembly and core models. The results showed comparable performance to an identical UO2 fueled simulation with the same configuration. The parameters simulated for composite and oxide fuel include the following: fuel to moderator ratio curves; energy dependent flux spectra; temperature coefficients for fuel and moderator; delayed neutron fractions; power peaking factors; axial and radial flux profiles in 2D and 3D; burnup; critical boron concentration; and shutdown margin. Overall, the neutronic parameters suggest that the transition from UO2 to composite in existing nuclear systems will not require significant changes in operating procedures or modifications to standards and regulations.

POBAL-S, the analysis and design of a high altitude airship. Final report, October 1972--March 1975. [For station keeping at an altitude of 21 km for 7 days; 500 W fuel cell power supply

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beemer, J.D.; Parsons, R.R.; Rueter, L.L.

1975-02-01

An engineering analysis and development effort has been executed to design a superpressure airship, POBAL-S, capable of station keeping at an altitude of 21 kilometers for a duration of 7 days while supporting a payload weighing 890 Newtons and requiring 500 watts of electrical power. A detailed parametric trade-off between various power sources and other design choices was performed. The computer program used to accomplish this analysis is described and many results are presented. The system concept which resulted was a fuel cell powered, propeller driven airship controlled by an on-board autopilot with basic commands telemetered from a ground controlmore » station. Design of the balloon, power train, gimbaled propeller assembly, and electronic/electrical systems is presented. Flight operations for launch and recovery are discussed.« less

Self-actuated nuclear reactor shutdown system using induction pump to facilitate sensing of core coolant temperature

DOEpatents

Sievers, Robert K.; Cooper, Martin H.; Tupper, Robert B.

1987-01-01

A self-actuated shutdown system incorporated into a reactivity control assembly in a nuclear reactor includes pumping means for creating an auxiliary downward flow of a portion of the heated coolant exiting from the fuel assemblies disposed adjacent to the control assembly. The shutdown system includes a hollow tubular member which extends through the outlet of the control assembly top nozzle so as to define an outer annular flow channel through the top nozzle outlet separate from an inner flow channel for primary coolant flow through the control assembly. Also, a latching mechanism is disposed in an inner duct of the control assembly and is operable for holding absorber bundles in a raised position in the control assembly and for releasing them to drop them into the core of the reactor for shutdown purposes. The latching mechanism has an inner flow passage extending between and in flow communication with the absorber bundles and the inner flow channel of the top nozzle for accommodating primary coolant flow upwardly through the control assembly. Also, an outer flow passage separate from the inner flow passage extends through the latching mechanism between and in flow communication with the inner duct and the outer flow channel of the top nozzle for accommodating inflow of a portion of the heated coolant from the adjacent fuel assemblies. The latching mechanism contains a magnetic material sensitive to temperature and operable to cause mating or latching together of the components of the latching mechanism when the temperature sensed is below a known temperature and unmating or unlatching thereof when the temperature sensed is above a given temperature. The temperature sensitive magnetic material is positioned in communication with the heated coolant flow through the outer flow passage for directly sensing the temperature thereof. Finally, the pumping means includes a jet induction pump nozzle and diffuser disposed adjacent the bottom nozzle of the control assembly and in flow communication with the inlet thereof. The pump nozzle is operable to create an upward driving flow of primary coolant through the pump diffuser and then to the absorber bundles. The upward driving flow of primary coolant, in turn, creates a suction head within the outer flow channel of the top nozzle and thereby an auxiliary downward flow of the heated coolant portion exiting from the upper end of the adjacent fuel assemblies through the outer flow channel to the pump nozzle via the outer flow passage of the latching mechanism and an annular space between the outer and inner spaced ducts of the control assembly housing. The temperature of the heated coolant exiting from the adjacent fuel assemblies can thereby be sensed directly by the temperature sensitive magnetic material in the latching mechanism.

78 FR 5773 - Notification of Proposed Production Activity, Generac Power Systems, Inc., Subzone 41J...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-28

... assemblies; oil/fuel filters; air/oil separation equipment; air filters/elements; catalytic converters... assemblies; AC line filters; dielectric items of paper/plastic; capacitors; circuit breakers; switching...

Tubular screen electrical connection support for solid oxide fuel cells

DOEpatents

Tomlins, Gregory W.; Jaszcar, Michael P.

2002-01-01

A solid oxide fuel assembly is made of fuel cells (16, 16', 18, 24, 24', 26), each having an outer interconnection layer (36) and an outer electrode (28), which are disposed next to each other with rolled, porous, hollow, electrically conducting metal mesh conductors (20, 20') between the fuel cells, connecting the fuel cells at least in series along columns (15, 15') and where there are no metal felt connections between any fuel cells.

Weathering Patterns of Ignitable Liquids with the Advanced Distillation Curve Method

PubMed Central

Bruno, Thomas J; Allen, Samuel

2013-01-01

One can take advantage of the striking similarity of ignitable liquid vaporization (or weathering) patterns and the separation observed during distillation to predict the composition of residual compounds in fire debris. This is done with the advanced distillation curve (ADC) metrology, which separates a complex fluid by distillation into fractions that are sampled, and for which thermodynamically consistent temperatures are measured at atmospheric pressure. The collected sample fractions can be analyzed by any method that is appropriate. Analytical methods we have applied include gas chromatography (with flame ionization, mass spectrometric and sulfur chemiluminescence detection), thin layer chromatography, FTIR, Karl Fischer coulombic titrimetry, refractometry, corrosivity analysis, neutron activation analysis and cold neutron prompt gamma activation analysis. We have applied this method on product streams such as finished fuels (gasoline, diesel fuels, aviation fuels, rocket propellants), crude oils (including a crude oil made from swine manure) and waste oils streams (used automotive and transformer oils). In this paper, we present results on a variety of ignitable liquids that are not commodity fuels, chosen from the Ignitable Liquids Reference Collection (ILRC). These measurements are assembled into a preliminary database. From this selection, we discuss the significance and forensic application of the temperature data grid and the composition explicit data channel of the ADC. PMID:26401423

Weathering Patterns of Ignitable Liquids with the Advanced Distillation Curve Method.

PubMed

Bruno, Thomas J; Allen, Samuel

2013-01-01

One can take advantage of the striking similarity of ignitable liquid vaporization (or weathering) patterns and the separation observed during distillation to predict the composition of residual compounds in fire debris. This is done with the advanced distillation curve (ADC) metrology, which separates a complex fluid by distillation into fractions that are sampled, and for which thermodynamically consistent temperatures are measured at atmospheric pressure. The collected sample fractions can be analyzed by any method that is appropriate. Analytical methods we have applied include gas chromatography (with flame ionization, mass spectrometric and sulfur chemiluminescence detection), thin layer chromatography, FTIR, Karl Fischer coulombic titrimetry, refractometry, corrosivity analysis, neutron activation analysis and cold neutron prompt gamma activation analysis. We have applied this method on product streams such as finished fuels (gasoline, diesel fuels, aviation fuels, rocket propellants), crude oils (including a crude oil made from swine manure) and waste oils streams (used automotive and transformer oils). In this paper, we present results on a variety of ignitable liquids that are not commodity fuels, chosen from the Ignitable Liquids Reference Collection (ILRC). These measurements are assembled into a preliminary database. From this selection, we discuss the significance and forensic application of the temperature data grid and the composition explicit data channel of the ADC.

75 FR 27491 - Airworthiness Directives; Pratt & Whitney Canada Corp. PW617F-E Turbofan Engines

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

.... Investigation showed that the Fuel Filter Bypass Valve poppet in the Fuel Oil Heat Exchanger (FOHE) on that... Fuel Filter Bypass Valve poppet in the FOHE on that engine had worn through the housing seat, allowing.... PW600-72-A66021 that introduced a new fuel Filter Bypass Valve Assembly with an improved design poppet...

Guidebook on LANDFIRE fuels data acquisition, critique, modification, maintenance, and model calibration

Treesearch

Richard D. Stratton

2009-01-01

With the advent of LANDFIRE fuels layers, an increasing number of specialists are using the data in a variety of fire modeling systems. However, a comprehensive guide on acquiring, critiquing, and editing (ACE) geospatial fuels data does not exist. This paper provides guidance on ACE as well as on assembling a geospatial fuels team, model calibration, and maintaining...

KSC-2009-2874

NASA Image and Video Library

2009-04-28

CAPE CANAVERAL, Fla. –– A variety of alternative fuel vehicles are driven around NASA's Kennedy Space Center in Florida in an effort to reduce gasoline consumption and conserve energy. These include compressed natural gas, bi-fuel, diesel fuel and flex fuel vehicles. Here they are on display at the NASA News Center. In the background is the Vehicle Assembly Building. Photo credit: NASA/Jim Grossmann

KSC-2009-2873

NASA Image and Video Library

2009-04-28

CAPE CANAVERAL, Fla. –– A variety of alternative fuel vehicles are driven around NASA's Kennedy Space Center in Florida in an effort to reduce gasoline consumption and conserve energy. These include compressed natural gas, bi-fuel, diesel fuel and flex fuel vehicles. Here they are on display at the NASA News Center. In the background is the Vehicle Assembly Building. Photo credit: NASA/Jim Grossmann

KSC-2009-2871

NASA Image and Video Library

2009-04-28

CAPE CANAVERAL, Fla. –– A variety of alternative fuel vehicles are driven around NASA's Kennedy Space Center in Florida in an effort to reduce gasoline consumption and conserve energy. These include compressed natural gas, bi-fuel, diesel fuel and flex fuel vehicles. Here they are on display at the NASA News Center. In the background is the Vehicle Assembly Building. Photo credit: NASA/Jim Grossmann

Temporal Control over Transient Chemical Systems using Structurally Diverse Chemical Fuels.

PubMed

Chen, Jack L-Y; Maiti, Subhabrata; Fortunati, Ilaria; Ferrante, Camilla; Prins, Leonard J

2017-08-25

The next generation of adaptive, intelligent chemical systems will rely on a continuous supply of energy to maintain the functional state. Such systems will require chemical methodology that provides precise control over the energy dissipation process, and thus, the lifetime of the transiently activated function. This manuscript reports on the use of structurally diverse chemical fuels to control the lifetime of two different systems under dissipative conditions: transient signal generation and the transient formation of self-assembled aggregates. The energy stored in the fuels is dissipated at different rates by an enzyme, which installs a dependence of the lifetime of the active system on the chemical structure of the fuel. In the case of transient signal generation, it is shown that different chemical fuels can be used to generate a vast range of signal profiles, allowing temporal control over two orders of magnitude. Regarding self-assembly under dissipative conditions, the ability to control the lifetime using different fuels turns out to be particularly important as stable aggregates are formed only at well-defined surfactant/fuel ratios, meaning that temporal control cannot be achieved by simply changing the fuel concentration. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Photography by KSC Space Shuttle Orbiter Enterprise mated to an external fuel tank and two solid

NASA Technical Reports Server (NTRS)

1980-01-01

Photography by KSC Space Shuttle Orbiter Enterprise mated to an external fuel tank and two solid rocket boosters on top of a Mobil Launcher Platform, undergoes fit and function checks at the launch site for the first Space Shuttle at Launch Complex 39's Pad A. The dummy Space Shuttle was assembled in the Vehicle Assembly Building and rolled out to the launch site on May 1 as part of an exercise to make certain shuttle elements are compatible with the Spaceport's assembly and launch facilities and ground support equipment, and help clear the way for the launch of the Space Shuttle Orbiter Columbia.

PHOTOGRAPHY BY KSC SPACE SHUTTLE ORBITER ENTERPRISE MATED TO AN EXTERNAL FUEL TANK AND TWO SOLID

NASA Technical Reports Server (NTRS)

1980-01-01

PHOTOGRAPHY BY KSC SPACE SHUTTLE ORBITER ENTERPRISE MATED TO AN EXTERNAL FUEL TANK AND TWO SOLID ROCKET BOOSTERS ON TOP OF A MOBIL LAUNCHER PLATFORM, UNDERGOES FIT AND FUNCTION CHECKS AT THE LAUNCH SITE FOR THE FIRST SPACE SHUTTLE AT LAUNCH COMPLEX 39'S PAD A. THE DUMMY SPACE SHUTTLE WAS ASSEMBLED IN THE VEHICLE ASSEMBLY BUILDING AND ROLLED OUT TO THE LAUNCH SITE ON MAY 1 AS PART OF AN EXERCISE TO MAKE CERTAIN SHUTTLE ELEMENTS ARE COMPATIBLE WITH THE SPACEPORT'S ASSEMBLY AND LAUNCH FACILITIES AND GROUND SUPPORT EQUIPMENT, AND HELP CLEAR THE WAY FOR THE LAUNCH OF THE SPACE SHUTTLE ORBITER COLUMBIA.

MELCOR Model of the Spent Fuel Pool of Fukushima Dai-ichi Unit 4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbajo, Juan J

2012-01-01

Unit 4 of the Fukushima Dai-ichi Nuclear Power Plant suffered a hydrogen explosion at 6:00 am on March 15, 2011, exactly 3.64 days after the earthquake hit the plant and the off-site power was lost. The earthquake occurred on March 11 at 2:47 pm. Since the reactor of this Unit 4 was defueled on November 29, 2010, and all its fuel was stored in the spent fuel pool (SFP4), it was first believed that the explosion was caused by hydrogen generated by the spent fuel, in particular, by the recently discharged core. The hypothetical scenario was: power was lost, coolingmore » to the SFP4 water was lost, pool water heated/boiled, water level decreased, fuel was uncovered, hot Zircaloy reacted with steam, hydrogen was generated and accumulated above the pool, and the explosion occurred. Recent analyses of the radioisotopes present in the water of the SFP4 and underwater video indicated that this scenario did not occur - the fuel in this pool was not damaged and was never uncovered the hydrogen of the explosion was apparently generated in Unit 3 and transported through exhaust ducts that shared the same chimney with Unit 4. This paper will try to answer the following questions: Could that hypothetical scenario in the SFP4 had occurred? Could the spent fuel in the SPF4 generate enough hydrogen to produce the explosion that occurred 3.64 days after the earthquake? Given the magnitude of the explosion, it was estimated that at least 150 kg of hydrogen had to be generated. As part of the investigations of this accident, MELCOR models of the SFP4 were prepared and a series of calculations were completed. The latest version of MELCOR, version 2.1 (Ref. 1), was employed in these calculations. The spent fuel pool option for BWR fuel was selected in MELCOR. The MELCOR model of the SFP4 consists of a total of 1535 fuel assemblies out of which 548 assemblies are from the core defueled on Nov. 29, 2010, 783 assemblies are older assemblies, and 204 are new/fresh assemblies. The total decay heat of the fuel in the pool was, at the time of the accident, 2.284 MWt, of which 1.872 MWt were from the 548 assemblies of the last core discharged and 0.412 MWt were from the older 783 assemblies. These decay heat values were calculated at Oak Ridge National Laboratory using the ORIGEN2.2 code (Ref. 2) - they agree with values reported elsewhere (Ref. 3). The pool dimensions are 9.9 m x 12.2 m x 11.8 m (height), and with the water level at 11.5 m, the pool volume is 1389 m3, of which only 1240 m3 is water, as some volume is taken by the fuel and by the fuel racks. The initial water temperature of the SFP4 was assumed to be 301 K. The fuel racks are made of an aluminum alloy but are modeled in MELCOR with stainless steel and B4C. MELCOR calculations were completed for different initial water levels: 11.5 m (pool almost full, water is only 0.3 m below the top rim), 4.4577 m (top of the racks), 4.2 m, and 4.026 m (top of the active fuel). A calculation was also completed for a rapid loss of water due to a leak at the bottom of the pool, with the fuel rapidly uncovered and oxidized in air. Results of these calculations are shown in the enclosed Table I. The calculation with the initial water level at 11.5 m (full pool) takes 11 days for the water to boil down to the top of the fuel racks, 11.5 days for the fuel to be uncovered, 14.65 days to generate 150 kg of hydrogen and 19 days for the pool to be completely dry. The calculation with the initial water level at 4.4577 m, takes 1.1 days to uncover the fuel and 4.17 days to generate 150 kg of hydrogen. The calculation with the initial water level at 4.02 m takes 3.63 days to generate 150 kg of hydrogen this is exactly the time when the actual explosion occurred in Unit 4. Finally, fuel oxidation in air after the pool drained the water in 20 minutes, generates only 10 kg of hydrogen this is because very little steam is available and Zircaloy (Zr) oxidation with the oxygen of the air does not generate hydrogen. MELCOR calculated water levels and hydrogen generated in the SFP4 as a function of time for initial water levels of 4.457 m, 4.2 m and 4.02 m are shown in Figs. 1 and 2. Water levels increase at the beginning due to the expansion of the water during the heat-up from 301 K to 373 K. Boiling occurs after the water temperature reaches 373 K. The total amount of hydrogen generated is ~2000 kg, this amount includes hydrogen generated from Zr, which is the largest amount (~1580 kg), from stainless steel (~360 kg), and from B4C (~60 kg). In theory, it is possible to generate up to 3.4 kg of hydrogen per assembly (from oxidation of Zr in the fuel cladding and box), or a total of 4,525 kg from the hot 1331 assemblies stored in the SFP4. The hydrogen generated from oxidation of steel and B4C will be additional. So the answers to the questions are YES according to these MELCOR calculations, enough hydrogen (150 kg) could be generated in the SFP4 3.64 days after the earthquake to produce ...« less

Active Interrogation for Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; Dougan, Arden

2015-11-05

The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE PAGES

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.; ...

2017-08-02

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Bays; W. Skerjanc; M. Pope

A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch averagemore » discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

New Autonomous Motors of Metal-Organic Framework (MOF) Powered by Reorganization of Self-Assembled Peptides at interfaces

PubMed Central

Ikezoe, Yasuhiro; Washino, Gosuke; Uemura, Takashi; Kitagawa, Susumu; Matsui, Hiroshi

2012-01-01

There have developed a variety of microsystems that harness energy and convert it to mechanical motion. Here we developed new autonomous biochemical motors by integrating metal-organic framework (MOF) and self-assembling peptides. MOF is applied as an energy-storing cell that assembles peptides inside nanoscale pores of the coordination framework. The robust assembling nature of peptides enables reconfiguring their assemblies at the water-MOF interface, which is converted to fuel energy. Re-organization of hydrophobic peptides could create the large surface tension gradient around the MOF and it efficiently powers the translation motion of MOF. As a comparison, the velocity of normalized by volume for the DPA-MOF particle is faster and the kinetic energy per the unit mass of fuel is more than twice as large as the one for previous gel motor systems. This demonstration opens the new application of MOF and reconfigurable molecular self-assembly and it may evolve into the smart autonomous motor that mimic bacteria to swim and harvest target chemicals by integrating recognition units. PMID:23104155

Emergency core cooling system

DOEpatents

Schenewerk, William E.; Glasgow, Lyle E.

1983-01-01

A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

Restartable High Power Gas Generator.

DTIC Science & Technology

1982-12-01

RELEASE; DISTRIBUTION UNLIMITED DTIC AERO PROPULSION LABORATORY " D T-C ) AIR FORCE WRIGHT AERONAUTICAL LABORATORIES AIR FORCE SYSTEMS COMMAND APR 2 3 133 C...INIFRV 4umat OVT CCESIONNO. 3 .PERCORMNG’ CAORGO NUMER 14. ATLE(ad S.##* CTPOfRCTOR GR N OERED M. G. Gants F33615-79-C-2004 9. PERFORMING ORGANIZATION...Assembly 11 3 Injector Internal Confiquration 12 4 Injector Assembly 15 5 Injector Housing le 6 Pintle 17 7 Core 18 8 Fuel Injection Rina iS 9 Fuel

Fuel cell cooler assembly and edge seal means therefor

DOEpatents

Breault, Richard D.; Roethlein, Richard J.; Congdon, Joseph V.

1980-01-01

A cooler assembly for a stack of fuel cells comprises a fibrous, porous coolant tube holder sandwiched between and bonded to at least one of a pair of gas impervious graphite plates. The tubes are disposed in channels which pass through the holder. The channels are as deep as the holder thickness, which is substantially the same as the outer diameter of the tubes. Gas seals along the edges of the holder parallel to the direction of the channels are gas impervious graphite strips.

BORAX V EXPONENTIAL EXPERIMENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirn, F.S.; Hagen, J.I.

1963-04-01

The cadmium ratio was measured in an exponential mockup of Borax V as a function of the void fraction. The extent of voids, simulated by lengths of closed polyethylene tubes, ranged from 0 to 40%. The corresponding cadmium ratios ranged from 6.1 to 4.6. The exponential was also used to determine the radial flux pattern across a Borax-type fuel assembly and the fine flux detail in and around fuel rods. For a normal loading the maximum-to-average power generation across an assembly was 1.24. (auth)

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, C.W.

1992-03-24

A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

Laser cutting apparatus for nuclear core fuel subassembly

DOEpatents

Walch, Allan P.; Caruolo, Antonio B.

1982-02-23

The object of the invention is to provide a system and apparatus which employs laser cutting to disassemble a nuclear core fuel subassembly. The apparatus includes a gantry frame (C) which straddles the core fuel subassembly (14), an x-carriage (22) travelling longitudinally above the frame which carries a focus head assembly (D) having a vertically moving carriage (46) and a laterally moving carriage (52), a system of laser beam transferring and focusing mirrors carried by the x-carriage and focusing head assembly, and a shroud follower (F) and longitudinal follower (G) for following the shape of shroud (14) to maintain a beam focal point (44) fixed upon the shroud surface for accurate cutting.

PEM/SPE fuel cell

DOEpatents

Grot, S.A.

1998-01-13

A PEM/SPE fuel cell is described including a membrane-electrode assembly (MEA) having a plurality of oriented filament embedded the face thereof for supporting the MEA and conducting current therefrom to contiguous electrode plates. 4 figs.

Fabrication of gas impervious edge seal for a bipolar gas distribution assembly for use in a fuel cell

DOEpatents

Kaufman, Arthur; Werth, John

1986-01-01

A bipolar gas reactant distribution assembly for use in a fuel cell is disclosed, the assembly having a solid edge seal to prevent leakage of gaseous reactants wherein a pair of porous plates are provided with peripheral slits generally parallel to, and spaced apart from two edges of the plate, the slit being filled with a solid, fusible, gas impervious edge sealing compound. The plates are assembled with opposite faces adjacent one another with a layer of a fusible sealant material therebetween the slits in the individual plates being approximately perpendicular to one another. The plates are bonded to each other by the simultaneous application of heat and pressure to cause a redistribution of the sealant into the pores of the adjacent plate surfaces and to cause the edge sealing compound to flow and impregnate the region of the plates adjacent the slits and comingle with the sealant layer material to form a continuous layer of sealant along the edges of the assembled plates.

Layer-by-layer self-assembly in the development of electrochemical energy conversion and storage devices from fuel cells to supercapacitors.

PubMed

Xiang, Yan; Lu, Shanfu; Jiang, San Ping

2012-11-07

As one of the most effective synthesis tools, layer-by-layer (LbL) self-assembly technology can provide a strong non-covalent integration and accurate assembly between homo- or hetero-phase compounds or oppositely charged polyelectrolytes, resulting in highly-ordered nanoscale structures or patterns with excellent functionalities and activities. It has been widely used in the developments of novel materials and nanostructures or patterns from nanotechnologies to medical fields. However, the application of LbL self-assembly in the development of highly efficient electrocatalysts, specific functionalized membranes for proton exchange membrane fuel cells (PEMFCs) and electrode materials for supercapacitors is a relatively new phenomenon. In this review, the application of LbL self-assembly in the development and synthesis of key materials of PEMFCs including polyelectrolyte multilayered proton-exchange membranes, methanol-blocking Nafion membranes, highly uniform and efficient Pt-based electrocatalysts, self-assembled polyelectrolyte functionalized carbon nanotubes (CNTs) and graphenes will be reviewed. The application of LbL self-assembly for the development of multilayer nanostructured materials for use in electrochemical supercapacitors will also be reviewed and discussed (250 references).

Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Majed, M.; Morback, G.; Wiman, P.

1995-09-01

The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give largemore » advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.« less

An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L

2009-01-01

There are a variety of reasons for quantifying plutonium (Pu) in spent fuel. Below, five motivations are listed: (1) To verify the Pu content of spent fuel without depending on unverified information from the facility, as requested by the IAEA ('independent verification'). New spent fuel measurement techniques have the potential to allow the IAEA to recover continuity of knowledge and to better detect diversion. (2) To assure regulators that all of the nuclear material of interest leaving a nuclear facility actually arrives at another nuclear facility ('shipper/receiver'). Given the large stockpile of nuclear fuel at reactor sites around the world,more » it is clear that in the coming decades, spent fuel will need to be moved to either reprocessing facilities or storage sites. Safeguarding this transportation is of significant interest. (3) To quantify the Pu in spent fuel that is not considered 'self-protecting.' Fuel is considered self-protecting by some regulatory bodies when the dose that the fuel emits is above a given level. If the fuel is not self-protecting, then the Pu content of the fuel needs to be determined and the Pu mass recorded in the facility's accounting system. This subject area is of particular interest to facilities that have research-reactor spent fuel or old light-water reactor (LWR) fuel. It is also of interest to regulators considering changing the level at which fuel is considered self-protecting. (4) To determine the input accountability value at an electrochemical processing facility. It is not expected that an electrochemical reprocessing facility will have an input accountability tank, as is typical in an aqueous reprocessing facility. As such, one possible means of determining the input accountability value is to measure the Pu content in the spent fuel that arrives at the facility. (5) To fully understand the composition of the fuel in order to efficiently and safely pack spent fuel into a long-term repository. The NDA of spent fuel can be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full assembly library for measurements in three different media: air, water and borated water. The 12 NDA techniques being researched are the following: Delayed Gamma, Delayed Neutrons, Differential Die-Away, Lead Slowing Down Spectrometer, Neutron Multiplicity, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Passive Neutron Albedo Reactivity, Self-integration Neutron Resonance Densitometry, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection.« less

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bradley K. Heath

2014-03-01

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show thatmore » fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.« less

Motor vehicle technology:Mobility for prosperity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1985-01-01

This book presents the papers given at a conference on internal combustion engines for vehicles. Topics considered at the conference included combustion chambers, the lubrication of turbocharged engines, oil filters, fuel consumption, traffic control, crashworthiness, brakes, acceleration, unleaded gasoline, methanol fuels, pressure drop, safety regulations, tire vibration, detergents, fuel economy, ceramics in engines, steels, catalytic converters, fuel additives, heat exchangers, pump systems, emissions control, fuel injection systems, noise pollution control, natural gas fuels, assembly plant productivity, aerodynamics, torsion, electronics, and automatic transmissions.

Modeling down-scattered neutron images from cryogenic fuel implosions at the National Ignition Facility

NASA Astrophysics Data System (ADS)

Raman, Kumar; Casey, Dan; Callahan, Debra; Clark, Dan; Fittinghoff, David; Grim, Gary; Hatchett, Steve; Hinkel, Denise; Jones, Ogden; Kritcher, Andrea; Seek, Scott; Suter, Larry; Merrill, Frank; Wilson, Doug

2016-10-01

In experiments with cryogenic deuterium-tritium (DT) fuel layers at the National Ignition Facility (NIF), an important technique for visualizing the stagnated fuel assembly is to image the 6-12 MeV neutrons created by scatters of the 14 MeV hotspot neutrons in the surrounding cold fuel. However, such down-scattered neutron images are difficult to interpret without a model of the fuel assembly, because of the nontrivial neutron kinematics involved in forming the images. For example, the dominant scattering modes are at angles other than forward scattering and the 14 MeV neutron fluence is not uniform. Therefore, the intensity patterns in these images usually do not correspond in a simple way to patterns in the fuel distribution, even for simple fuel distributions. We describe our efforts to model synthetic images from ICF design simulations with data from the National Ignition Campaign and after. We discuss the insight this gives, both to understand how well the models are predicting fuel asymmetries and to inform how to optimize the diagnostic for the types of fuel distributions being predicted. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Comparison Of 252Cf Time Correlated Induced Fisssion With AmLi Induced Fission On Fresh MTR Research Reactor Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Joshi, Jay Prakash

The effective application of international safeguards to research reactors requires verification of spent fuel as well as fresh fuel. To accomplish this goal various nondestructive and destructive assay techniques have been developed in the US and around the world. The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) system developed at Los Alamos National Laboratory (LANL) combining both neutron and gamma measurement capabilities. Since spent fuel assemblies are stored in water, the system was designed to be watertight to facilitate underwater measurements by inspectors. The AEFC is comprised of six 3He detectors as well as a shielded andmore » collimated ion chamber. The 3He detectors are used for active and passive neutron coincidence counting while the ion chamber is used for gross gamma counting. Active coincidence measurement data is used to measure residual fissile mass, whereas the passive coincidence measurement data along with passive gamma measurement can provide information about burnup, cooling time, and initial enrichment. In the past, most of the active interrogation systems along with the AEFC used an AmLi neutron interrogation source. Owing to the difficulty in obtaining an AmLi source, a 252Cf spontaneous fission (SF) source was used during a 2014 field trail in Uzbekistan as an alternative. In this study, experiments were performed to calibrate the AEFC instrument and compare use of the 252Cf spontaneous fission source and the AmLi (α,n) neutron emission source. The 252Cf source spontaneously emits bursts of time-correlated prompt fission neutrons that thermalize in the water and induce fission in the fuel assembly. The induced fission (IF) neutrons are also time correlated resulting in more correlated neutron detections inside the 3He detector, which helps reduce the statistical errors in doubles when using the 252Cf interrogation source instead of the AmLi source. In this work, two MTR fuel assemblies varying both in size and number of fuel plates were measured using 252Cf and AmLi active interrogation sources. This paper analyzes time correlated induced fission (TCIF) from fresh MTR fuel assemblies due to 252Cf and AmLi active interrogation sources.« less

Integration of Rotor Aerodynamic Optimization with the Conceptual Design of a Large Civil Tiltrotor

DTIC Science & Technology

2010-01-01

Rotor MCP Maximum Continuous Power MRP Maximum Rated Power (take-off power) NDARC NASA Design and Analysis of Rotorcraft OEI One Engine Inoperative...OGE Out of Ground Effect SFC Specific Fuel Consumption SNI Simultaneous Non-Interfering approach STOL Short Takeoff and Landing VTOL Vertical...that are assembled into a complete aircraft model. NDARC is designed for high computational efficiency. Performance is calculated with physics- based

Improving Legacy Aircraft Systems Through Condition-Based Maintenance: An H-60 Case Study

DTIC Science & Technology

2014-09-01

level functions. These decompositions are equivalent to a detailed design effort in systems engineering. NAMPSOPs have a common architectural structure...Assembly Power Available Spindle Cables No.1 Engine Load Demand Spindle Control Cables Engine Pneumatic Starters Auxiliary Power Unit IRCM FLIR Mission...Analysis Fuel System Main Rotor Head Main Module Main Gear Box Radiator Engine Output Shaft Auxiliary Power Unit Flight Control Cables Tail Landing

Apparatus for testing high pressure injector elements

NASA Technical Reports Server (NTRS)

Myers, William Neill (Inventor); Scott, Ewell M. (Inventor); Forbes, John C. (Inventor); Shadoan, Michael D. (Inventor)

1995-01-01

An apparatus for testing and evaluating the spray pattern of high pressure fuel injector elements for use in supplying fuel to combustion engines is presented. Prior art fuel injector elements were normally tested by use of low pressure apparatuses which did not provide a purge to prevent mist from obscuring the injector element or to prevent frosting of the view windows; could utilize only one fluid during each test; and had their viewing ports positioned one hundred eighty (180 deg) apart, thus preventing optimum use of laser diagnostics. The high pressure fluid injector test apparatus includes an upper hub, an upper weldment or housing, a first clamp and stud/nut assembly for securing the upper hub to the upper weldment, a standoff assembly within the upper weldment, a pair of window housings having view glasses within the upper weldment, an injector block assembly and purge plate within the upper weldment for holding an injector element to be tested and evaluated, a lower weldment or housing, a second clamp and stud/nut assembly for securing the lower weldment to the upper hub, a third clamp and stud/nut assembly for securing the lower hub to the lower weldment, mechanisms for introducing fluid under high pressure for testing an injector element, and mechanisms for purging the apparatus to prevent frosting of view glasses within the window housings and to permit unobstructed viewing of the injector element.

Apparatus for testing high pressure injector elements

NASA Technical Reports Server (NTRS)

Myers, William Neill (Inventor); Scott, Ewell M. (Inventor); Forbes, John C. (Inventor); Shadoan, Michael D. (Inventor)

1993-01-01

An apparatus for testing and evaluating the spray pattern of high pressure fuel injector elements for use in supplying fuel to combustion engines is presented. Prior art fuel injector elements were normally tested by use of low pressure apparatuses which did not provide a purge to prevent mist from obscuring the injector element or to prevent frosting of the view windows; could utilize only one fluid during each test; and had their viewing ports positioned one hundred eighty (180 deg) apart, thus preventing optimum use of laser diagnostics. The high pressure fluid injector test apparatus includes an upper hub, an upper weldment or housing, a first clamp and stud/nut assembly for securing the upper hub to the upper weldment, a standoff assembly within the upper weldment, a pair of window housings having view glasses within the upper weldment, an injector block assembly and purge plate within the upper weldment for holding an injector element to be tested and evaluated, a lower weldment or housing, a second clamp and stud/nut assembly for securing the lower weldment to the upper weldment, a lower hub, a third clamp and stud/nut assembly for securing the lower hub to the lower weldment, mechanisms for introducing fluid under high pressure for testing an injector element, and mechanisms for purging the apparatus to prevent frosting of view glasses within the window housings and to permit unobstructed viewing of the injector element.