Fuel Lubricity Impact on Shipboard Engine and Fuel Systems and Sensitivity of U.S. Navy Diesel Engines to Low-Sulfur Diesel Fuel

DTIC Science & Technology

2011-06-30

load fuel and operated with a dummy injector to make sure the system was clean. The rig was de -fueled and a fresh charge of 2000-gram fuel was added...the rocker arm on the injector. The rocker arm contact was repositioned when it was noted it was hitting the injector off-center, and it was felt...going up. Figure B6. DD 149 Unit Injector with Diesel Fuel and Centered Rocker Arm Figure B7. Wear Rate Deviation Attributed to Head

Submission of FeCrAl Feedstock for Support of AFC ATR-2 Irradiations

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

Field, Kevin G.; Barrett, Kristine E.; Sun, Zhiqian

The Advanced Test Reactor (ATR) is currently being used to test accident tolerant fuel (ATF) forms destined for commercial nuclear power plant deployment. One irradiation program using the ATR for ATF concepts, Accident Tolerant Fuel-2 (ATF-2), is a water loop irradiation test using miniaturized fuel pins as test articles. This complicated testing configuration requires a series of pre-test experiments and verification including a flowing loop autoclave test and a sensor qualification test (SQT) prior to full test train deployment within the ATR. In support of the ATF-2 irradiation program, Oak Ridge National Laboratory (ORNL) has supplied two different Generation IImore » FeCrAl alloys in rod stock form to Idaho National Laboratory (INL). These rods will be machined into dummy pins for deployment in the autoclave test and SQT. Post-test analysis of the dummy pins will provide initial insight into the performance of Generation II FeCrAl alloys in the ATF-2 irradiation experiment as well as within a commercial nuclear reactor.« less

Validation of Finite Element Crash Test Dummy Models for Predicting Orion Crew Member Injuries During a Simulated Vehicle Landing

NASA Technical Reports Server (NTRS)

Tabiei, Al; Lawrence, Charles; Fasanella, Edwin L.

2009-01-01

A series of crash tests were conducted with dummies during simulated Orion crew module landings at the Wright-Patterson Air Force Base. These tests consisted of several crew configurations with and without astronaut suits. Some test results were collected and are presented. In addition, finite element models of the tests were developed and are presented. The finite element models were validated using the experimental data, and the test responses were compared with the computed results. Occupant crash data, such as forces, moments, and accelerations, were collected from the simulations and compared with injury criteria to assess occupant survivability and injury. Some of the injury criteria published in the literature is summarized for completeness. These criteria were used to determine potential injury during crew impact events.

Evaluating the Effectiveness of Various Blast Loading Descriptors as Occupant Injury Predictors for Underbody Blast Events

DTIC Science & Technology

2014-01-09

of Hybrid III ATD LSDYNA model with FTSS v7.1.6 finite element dummy 6 Unclassified: Distribution Statement A. Approved for public release...descriptors as occupant injury predictors for underbody blast events Recording injury metrics Response from the dummy especially pelvic acceleration and...Ciip(H&ad CG,2) "’"’ "-......--------, I Max : 122.669 @59.81 7!; Time, ms Pelvic Z acceleration, g I I Clip: -4.75737 Ts:97.4138 Te: 104.414

A Finite Element Model of the THOR-K Dummy for Aerospace and Aircraft Impact Simulations

NASA Technical Reports Server (NTRS)

Putnam, Jacob; Untaroiu, Costin D.; Somers, Jeffrey T.; Pellettiere, Joseph

2013-01-01

1) Update and Improve the THOR Finite Element (FE) model to specifications of the latest mod kit (THOR-K). 2) Evaluate the kinematic and kinetic response of the FE model in frontal, spinal, and lateral impact loading conditions.

Controlled Impact Demonstration instrumented test dummies installed in plane

NASA Technical Reports Server (NTRS)

1984-01-01

In this photograph are seen some of dummies in the passenger cabin of the B-720 aircraft. NASA Langley Research Center instrumented a large portion of the aircraft and the dummies for loads in a crashworthiness research program. In 1984 NASA Dryden Flight Research Facility and the Federal Aviation Adimistration (FAA) teamed-up in a unique flight experiment called the Controlled Impact Demonstration (CID). The test involved crashing a Boeing 720 aircraft with four JT3C-7 engines burning a mixture of standard fuel with an additive called Anti-misting Kerosene (AMK) designed to supress fire. In a typical aircraft crash, fuel spilled from ruptured fuel tanks forms a fine mist that can be ignited by a number of sources at the crash site. In 1984 the NASA Dryden Flight Research Facility (after 1994 a full-fledged Center again) and the Federal Aviation Administration (FAA) teamed-up in a unique flight experiment called the Controlled Impact Demonstration (CID), to test crash a Boeing 720 aircraft using standard fuel with an additive designed to supress fire. The additive, FM-9, a high-molecular-weight long-chain polymer, when blended with Jet-A fuel had demonstrated the capability to inhibit ignition and flame propagation of the released fuel in simulated crash tests. This anti-misting kerosene (AMK) cannot be introduced directly into a gas turbine engine due to several possible problems such as clogging of filters. The AMK must be restored to almost Jet-A before being introduced into the engine for burning. This restoration is called 'degradation' and was accomplished on the B-720 using a device called a 'degrader.' Each of the four Pratt & Whitney JT3C-7 engines had a 'degrader' built and installed by General Electric (GE) to break down and return the AMK to near Jet-A quality. In addition to the AMK research the NASA Langley Research Center was involved in a structural loads measurement experiment, which included having instrumented dummies filling the seats in the passenger compartment. Before the final flight on December 1, 1984, more than four years of effort passed trying to set-up final impact conditions considered survivable by the FAA. During those years while 14 flights with crews were flown the following major efforts were underway: NASA Dryden developed the remote piloting techniques necessary for the B-720 to fly as a drone aircraft; General Electric installed and tested four degraders (one on each engine); and the FAA refined AMK (blending, testing, and fueling a full-size aircraft). The 15 flights had 15 takeoffs, 14 landings and a larger number of approaches to about 150 feet above the prepared crash site under remote control. These flight were used to introduce AMK one step at a time into some of the fuel tanks and engines while monitoring the performance of the engines. On the final flight (No. 15) with no crew, all fuel tanks were filled with a total of 76,000 pounds of AMK and the remotely-piloted aircraft landed on Rogers Dry Lakebed in an area prepared with posts to test the effectiveness of the AMK in a controlled impact. The CID, which some wags called the Crash in the Desert, was spectacular with a large fireball enveloping and burning the B-720 aircraft. From the standpoint of AMK the test was a major set-back, but for NASA Langley, the data collected on crashworthiness was deemed successful and just as important.

Modeling and validation of a detailed FE viscoelastic lumbar spine model for vehicle occupant dummies.

PubMed

Amiri, Sorosh; Naserkhaki, Sadegh; Parnianpour, Mohamad

2018-06-19

The dummies currently used for predicting vehicle occupant response during frontal crashes or whole-body vibration provide insufficient information about spinal loads. Although they aptly approximate upper-body rotations in different loading scenarios, they overlook spinal loads, which are crucial to injury assessment. This paper aims to develop a modified dummy finite element (FE) model with a detailed viscoelastic lumbar spine. This model has been developed and validated against in-vitro and in-silico data under different loading conditions, and its predicted ranges of motion (RoM) and intradiscal pressure (IDP) maintain close correspondence with the in-vitro data. The dominant frequency of the model was f = 8.92 Hz, which was close to previous results. In the relaxation test, a force reduction of up to 21% was obtained, showing high agreement in force relaxation during the in-vitro test. The FE lumbar spine model was placed in the HYBRID III test dummy and aligned in a seated position based on available MRI data. Under two impulsive acceleration loadings in flexion and lateral directions with a peak acceleration of 60 m/s 2 , flexion responses of the modified and original dummies were close (RoMs of 29.1° and 29.6°, respectively), though not in lateral bending (RoMs of 34.1° and 15.6°, respectively), where the modified dummy was more flexible than the original. By reconstructing a real frontal crash, it was found that the modified dummy provided a 10% reduction in the Head Injury Criterion (HIC). Other than the more realistic behavior of this modified dummy, its capability of approximating lumbar loads and risk of lumbar spine injuries in vehicle crashes or whole-body vibration is of great importance. Copyright © 2018 Elsevier Ltd. All rights reserved.

Development of a curved pipe capability for the NASTRAN finite element program

NASA Technical Reports Server (NTRS)

Jeter, J. W., Jr.

1977-01-01

A curved pipe element capability for the NASTRAN structural analysis program is developed using the NASTRAN dummy element feature. A description is given of the theory involved in the subroutines which describe stiffness, mass, thermal and enforced deformation loads, and force and stress recovery for the curved pipe element. Incorporation of these subroutines into NASTRAN is discussed. Test problems are proposed. Instructions on use of the new element capability are provided.

The Influences of Arm Resist Motion on a CAR Crash Test Using Hybrid III Dummy with Human-Like Arm

NASA Astrophysics Data System (ADS)

Kim, Yongchul; Youm, Youngil; Bae, Hanil; Choi, Hyeonki

Safety of the occupant during the crash is very essential design element. Many researches have been investigated in reducing the fatal injury of occupant. They are focusing on the development of a dummy in order to obtain the real human-like motion. However, they have not considered the arm resist motion during the car accident. In this study, we would like to suggest the importance of the reactive force of the arm in a car crash. The influences of reactive force acting on the human upper extremity were investigated using the impedance experimental method with lumped mass model of hand system and a Hybrid III dummy with human-like arm. Impedance parameters (e.g. inertia, spring constant and damping coefficient) of the elbow joint in maximum activation level were measured by free oscillation test using single axis robot. The results showed that without seat belt, the reactive force of human arm reduced the head, chest, and femur injury, and the flexion moment of the neck is higher than that of the conventional dummy.

Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

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

Gryzinski, M.A.; Wielgosz, M.

The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel ismore » additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with self-cooling system (enhanced comparing to the cooling properties inherent with regular rector pool water flows), -proper cooling conditions can be ensured by an appropriate water flow, so the resistance to flow has to be optimised, -the requirement of the minimum resistance to water flow leads to the openwork design of the fuel element separator, which, on the other hand, has to be strong enough to ensure the needed strength for mechanical load due to the fuel weight and forces associated with the water flow, -the possibility of changing beam and flux qualities by rotating the converter or repositioning the converter plates by moving or replacing with another materials. In order to minimize the neutron activation of the fuel in the converter, the possibility was predicted to remove the converter and to replace it with an aluminium dummy for the time when the beam at the channel H2 is not used. This means that both, the converter and the dummy, have to be easily removable from the converter socket. There has to be also the place in the water pool, near the research stand or in technological pool, where the converter can be safely stored (this place have to be proper for operation with plates i.e. changing amount of plates). Thermal and neutron load of the fuel plates in the converter will be inhomogeneous. In order to equalize these loads, the converter should be designed in such way that it would be possible to change the order of fuel plates. Moreover replacing the amount of the plates gives the opportunity to obtain different fluxes of neutrons (quantitatively and qualitatively i.e. energetically). The project of the converter is based on Monte Carlo calculation concerning neutron production and on Computational Fluid Dynamics (CFD) i.e. modelling of converter for thermodynamical aspects. (authors)« less

The Influence of Neck Muscle Activation on Head and Neck Injuries of Occupants in Frontal Impacts.

PubMed

Li, Fan; Lu, Ronggui; Hu, Wei; Li, Honggeng; Hu, Shiping; Hu, Jiangzhong; Wang, Haibin; Xie, He

2018-01-01

The aim of the present paper was to study the influence of neck muscle activation on head and neck injuries of vehicle occupants in frontal impacts. A mixed dummy-human finite element model was developed to simulate a frontal impact. The head-neck part of a Hybrid III dummy model was replaced by a well-validated head-neck FE model with passive and active muscle characteristics. The mixed dummy-human FE model was validated by 15 G frontal volunteer tests conducted in the Naval Biodynamics Laboratory. The effects of neck muscle activation on the head dynamic responses and neck injuries of occupants in three frontal impact intensities, low speed (10 km/h), medium speed (30 km/h), and high speed (50 km/h), were studied. The results showed that the mixed dummy-human FE model has good biofidelity. The activation of neck muscles can not only lower the head resultant acceleration under different impact intensities and the head angular acceleration in medium- and high-speed impacts, thereby reducing the risks of head injury, but also protect the neck from injury in low-speed impacts.

Estimation of the auto frequency response function at unexcited points using dummy masses

NASA Astrophysics Data System (ADS)

Hosoya, Naoki; Yaginuma, Shinji; Onodera, Hiroshi; Yoshimura, Takuya

2015-02-01

If structures with complex shapes have space limitations, vibration tests using an exciter or impact hammer for the excitation are difficult. Although measuring the auto frequency response function at an unexcited point may not be practical via a vibration test, it can be obtained by assuming that the inertia acting on a dummy mass is an external force on the target structure upon exciting a different excitation point. We propose a method to estimate the auto frequency response functions at unexcited points by attaching a small mass (dummy mass), which is comparable to the accelerometer mass. The validity of the proposed method is demonstrated by comparing the auto frequency response functions estimated at unexcited points in a beam structure to those obtained from numerical simulations. We also consider random measurement errors by finite element analysis and vibration tests, but not bias errors. Additionally, the applicability of the proposed method is demonstrated by applying it to estimate the auto frequency response function of the lower arm in a car suspension.

Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

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

Brower, Jeffrey O.; Glazoff, Michael V.; Eiden, Thomas J.

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core andmore » inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping and pitting degradation on the YA-M fuel elements. In the case of scalloping (horseshoeing) a surprising similarity of that defect to those appearing on aluminum plate rolled in over-lubrication conditions, were established. In turn, this made us think that the principal feature responsible for the appearance of these defects, was horizontal cuts in the beryllium reflector block created to arrest the propagation of large vertical crack(s) in Be in PALM cycles with higher overall fluence. This assumption was fully confirmed by the results of thermo-hydraulic simulations. The neutronics data for these modeling experiments were provided using advanced irradiation simulations (MCNP, HELIOS). In the case of pitting erosion the following corrective measures were proposed based upon the results of JMatPro v.8.2 modeling (TTT- and CCT-diagrams): change the fabrication process by adding blister anneal before program anneal, immediately after cold rolling of AA6061plate. This step will allow achieving complete recrystallization, eliminating of strengthening due to metastable precipitates, and reduce the possibility of forming sharp microstructural features upon the surface.« less

Anthropometric specifications, development, and evaluation of EvaRID--a 50th percentile female rear impact finite element dummy model.

PubMed

Carlsson, Anna; Chang, Fred; Lemmen, Paul; Kullgren, Anders; Schmitt, Kai-Uwe; Linder, Astrid; Svensson, Mats Y

2014-01-01

Whiplash-associated disorders (WADs), or whiplash injuries, due to low-severity vehicle crashes are of great concern in motorized countries and it is well established that the risk of such injuries is higher for females than for males, even in similar crash conditions. Recent protective systems have been shown to be more beneficial for males than for females. Hence, there is a need for improved tools to address female WAD prevention when developing and evaluating the performance of whiplash protection systems. The objective of this study is to develop and evaluate a finite element model of a 50th percentile female rear impact crash test dummy. The anthropometry of the 50th percentile female was specified based on literature data. The model, called EvaRID (female rear impact dummy), was based on the same design concept as the existing 50th percentile male rear impact dummy, the BioRID II. A scaling approach was developed and the first version, EvaRID V1.0, was implemented. Its dynamic response was compared to female volunteer data from rear impact sled tests. The EvaRID V1.0 model and the volunteer tests compared well until ∼250 ms of the head and T1 forward accelerations and rearward linear displacements and of the head rearward angular displacement. Markedly less T1 rearward angular displacement was found for the EvaRID model compared to the female volunteers. Similar results were received for the BioRID II model when comparing simulated responses with experimental data under volunteer loading conditions. The results indicate that the biofidelity of the EvaRID V1.0 and BioRID II FE models have limitations, predominantly in the T1 rearward angular displacement, at low velocity changes (7 km/h). The BioRID II model was validated against dummy test results in a loading range close to consumer test conditions (EuroNCAP) and lower severity levels of volunteer testing were not considered. The EvaRID dummy model demonstrated the potential of becoming a valuable tool when evaluating and developing seats and whiplash protection systems. However, updates of the joint stiffness will be required to provide better correlation at lower load levels. Moreover, the seated posture, curvature of the spine, and head position of 50th percentile female occupants needs to be established and implemented in future models.

Optimizing the passenger air bag of an adaptive restraint system for multiple size occupants.

PubMed

Bai, Zhonghao; Jiang, Binhui; Zhu, Feng; Cao, Libo

2014-01-01

The development of the adaptive occupant restraint system (AORS) has led to an innovative way to optimize such systems for multiple size occupants. An AORS consists of multiple units such as adaptive air bags, seat belts, etc. During a collision, as a supplemental protective device, air bags can provide constraint force and play a role in dissipating the crash energy of the occupants' head and thorax. This article presents an investigation into an adaptive passenger air bag (PAB). The purpose of this study is to develop a base shape of a PAB for different size occupants using an optimization method. Four typical base shapes of a PAB were designed based on geometric data on the passenger side. Then 4 PAB finite element (FE) models and a validated sled with different size dummy models were developed in MADYMO (TNO, Rijswijk, The Netherlands) to conduct the optimization to obtain the best baseline PAB that would be used in the AORS. The objective functions-that is, the minimum total probability of injuries (∑Pcomb) of the 5th percentile female and 50th and 95th percentile male dummies-were adopted to evaluate the optimal configurations. The injury probability (Pcomb) for each dummy was adopted from the U.S. New Car Assessment Program (US-NCAP). The parameters of the AORS were first optimized for different types of PAB base shapes in a frontal impact. Then, contact time duration and force between the PAB and dummy head/chest were optimized by adjusting the parameters of the PAB, such as the number and position of tethers, lower the Pcomb of the 95th percentile male dummy. According to the optimization results, 4 typical PABs could provide effective protection to 5th and 50th percentile dummies. However, due to the heavy and large torsos of the 95th percentile occupants, the current occupant restraint system does not demonstrate satisfactory protective function, particularly for the thorax.

FY2017 Pilot Project Plan for the Nuclear Energy Knowledge and Validation Center Initiative

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

Ren, Weiju

To prepare for technical development of computational code validation under the Nuclear Energy Knowledge and Validation Center (NEKVAC) initiative, several meetings were held by a group of experts of the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory (ORNL) to develop requirements of, and formulate a structure for, a transient fuel database through leveraging existing resources. It was concluded in discussions of these meetings that a pilot project is needed to address the most fundamental issues that can generate immediate stimulus to near-future validation developments as well as long-lasting benefits to NEKVAC operation. The present project is proposedmore » based on the consensus of these discussions. Analysis of common scenarios in code validation indicates that the incapability of acquiring satisfactory validation data is often a showstopper that must first be tackled before any confident validation developments can be carried out. Validation data are usually found scattered in different places most likely with interrelationships among the data not well documented, incomplete with information for some parameters missing, nonexistent, or unrealistic to experimentally generate. Furthermore, with very different technical backgrounds, the modeler, the experimentalist, and the knowledgebase developer that must be involved in validation data development often cannot communicate effectively without a data package template that is representative of the data structure for the information domain of interest to the desired code validation. This pilot project is proposed to use the legendary TREAT Experiments Database to provide core elements for creating an ideal validation data package. Data gaps and missing data interrelationships will be identified from these core elements. All the identified missing elements will then be filled in with experimental data if available from other existing sources or with dummy data if nonexistent. The resulting hybrid validation data package (composed of experimental and dummy data) will provide a clear and complete instance delineating the structure of the desired validation data and enabling effective communication among the modeler, the experimentalist, and the knowledgebase developer. With a good common understanding of the desired data structure by the three parties of subject matter experts, further existing data hunting will be effectively conducted, new experimental data generation will be realistically pursued, knowledgebase schema will be practically designed; and code validation will be confidently planned.« less

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.

76 FR 31860 - Anthropomorphic Test Devices; Hybrid III Test Dummy, ES-2re Side Impact Crash Test Dummy

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-02

... [Docket No. NHTSA-2010-0146] RIN 2127-AK64 Anthropomorphic Test Devices; Hybrid III Test Dummy, ES-2re Side Impact Crash Test Dummy AGENCY: National Highway Traffic Safety Administration (NHTSA), Department..., 2008, concerning a 50th percentile adult male side crash test dummy called the ``ES-2re'' test dummy...

Placing three-dimensional isoparametric elements into NASTRAN. [alterations in matrix assembly to simplify generation of higher order elements

NASA Technical Reports Server (NTRS)

Newman, M. B.; Filstrup, A. W.

1973-01-01

Linear (8 node), parabolic (20 node), cubic (32 node) and mixed (some edges linear, some parabolic and some cubic) have been inserted into NASTRAN, level 15.1. First the dummy element feature was used to check out the stiffness matrix generation routines for the linear element in NASTRAN. Then, the necessary modules of NASTRAN were modified to include the new family of elements. The matrix assembly was changed so that the stiffness matrix of each isoparametric element is only generated once as the time to generate these higher order elements tends to be much longer than the other elements in NASTRAN. This paper presents some of the experiences and difficulties of inserting a new element or family of elements into NASTRAN.

A new mathematical neck model for a low-velocity rear-end impact dummy: evaluation of components influencing head kinematics.

PubMed

Linder, A

2000-03-01

A mathematical model of a new rear-end impact dummy neck was implemented using MADYMO. The main goal was to design a model with a human-like response of the first extension motion in the crash event. The new dummy neck was modelled as a series of rigid bodies (representing the seven cervical vertebrae and the uppermost thoracic element, T1) connected by pin joints, and supplemented by two muscle substitutes. The joints had non-linear stiffness characteristics and the muscle elements possessed both elastic stiffness and damping properties. The new model was compared with two neck models with the same number of vertebrae, but without muscle substitutes. The properties of the muscle substitutes and the need of these were evaluated by using three different modified neck models. The motion of T1 in the simulations was prescribed using displacement data obtained from volunteer tests. In a sensitivity analysis of the mathematical model the influence of different factors on the head-neck kinematics was evaluated. The neck model was validated against kinematics data from volunteer tests: linear displacement, angular displacement, and acceleration of the head relative to the upper torso at 7 km/h velocity change. The response of the new model was within the corridor of the volunteer tests for the main part of the time history plot. This study showed that a combination of elastic stiffness and damping in the muscle substitutes, together with a non-linear joint stiffness, resulted in a head-neck response similar to human volunteers, and superior to that of other tested neck models.

Full-face motorcycle helmet protection from facial impacts: an investigation using THOR dummy impacts and SIMon finite element head model.

PubMed

Whyte, Thomas; Gibson, Tom; Eager, David; Milthorpe, Bruce

2017-06-01

Facial impacts are both common and injurious for helmeted motorcyclists who crash; however, there is no facial impact requirement in major motorcycle helmet standards. This study examined the effect of full-face motorcycle helmet protection on brain injury risk in facial impacts using a test device with biofidelic head and neck motion. A preliminary investigation of energy absorbing foam in the helmet chin bar was carried out. Flat-faced rigid pendulum impacts were performed on a THOR dummy in an unprotected (no helmet) and protected mode (two full-face helmet conditions). The head responses of the dummy were input into the simulated injury monitor finite element head model to analyse the risk of brain injury in these impacts. Full-face helmet protection provides a significant reduction in brain injury risk in facial impacts at increasing impact speeds compared with an unprotected rider (p<0.05). The effect of low-density crushable foam added to the chin bar could not be distinguished from an unpadded chin bar impact. Despite the lack of an impact attenuation requirement for the face, full-face helmets do provide a reduction in head injury risk to the wearer in facial impacts. The specific helmet design factors that influence head injury risk in facial impacts need further investigation if improved protection for helmeted motorcyclists is to be achieved. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

Evaluation of a metal shear web selectively reinforced with filamentary composites for space shuttle application. Phase 2: summary report: Shear web component fabrication

NASA Technical Reports Server (NTRS)

Laakso, J. H.; Smith, D. D.; Zimmerman, D. K.

1973-01-01

The fabrication of two shear web test elements and three large scale shear web test components are reported. In addition, the fabrication of test fixtures for the elements and components is described. The center-loaded beam test fixtures were configured to have a test side and a dummy or permanent side. The test fixtures were fabricated from standard extruded aluminum sections and plates and were designed to be reuseable.

Mechanical qualification of the support structure for MQXF, the Nb 3Sn low-β quadrupole for the high luminosity LHC

DOE PAGES

Juchno, M.; Ambrosio, G.; Anerella, M.; ...

2016-01-26

Within the scope of the High Luminosity LHC project, the collaboration between CERN and U.S. LARP is developing new low-β quadrupoles using the Nb 3Sn superconducting technology for the upgrade of the LHC interaction regions. The magnet support structure of the first short model was designed and two units were fabricated and tested at CERN and at LBNL. The structure provides the preload to the collars-coils subassembly by an arrangement of outer aluminum shells pre-tensioned with water-pressurized bladders. For the mechanical qualification of the structure and the assembly procedure, superconducting coils were replaced with solid aluminum “dummy coils”, the structuremore » was preloaded at room temperature, and then cooled-down to 77 K. Mechanical behavior of the magnet structure was monitored with the use of strain gauges installed on the aluminum shells, the dummy coils and the axial preload system. As a result, this paper reports on the outcome of the assembly and the cool-down tests with dummy coils, which were performed at CERN and at LBNL, and presents the strain gauge measurements compared to the 3D finite element model predictions.« less

The Experimental Measurement of Local and Bulk Oxygen Transport Resistances in the Catalyst Layer of Proton Exchange Membrane Fuel Cells.

PubMed

Wang, Chao; Cheng, Xiaojing; Lu, Jiabin; Shen, Shuiyun; Yan, Xiaohui; Yin, Jiewei; Wei, Guanghua; Zhang, Junliang

2017-12-07

Remarkable progress has been made in reducing the cathodic Pt loading of PEMFCs; however, a huge performance loss appears at high current densities, indicating the existence of a large oxygen transport resistance associated with the ultralow Pt loading catalyst layer. To reduce the Pt loading without sacrificing cell performance, it is essential to illuminate the oxygen transport mechanism in the catalyst layer. Toward this goal, an experimental approach to measure the oxygen transport resistance in catalyst layers is proposed and realized for the first time in this study. The measuring approach involves a dual-layer catalyst layer design, which consists of a dummy catalyst layer and a practical catalyst layer, followed by changing the thickness of dummy layer to respectively quantify the local and bulk resistances via limiting current measurements combined with linear extrapolation. The experimental results clearly reveal that the local resistance dominates the total resistance in the catalyst layer.

Factors that influence chest injuries in rollovers.

PubMed

Digges, Kennerly; Eigen, Ana; Tahan, Fadi; Grzebieta, Raphael

2014-01-01

The design of countermeasures to reduce serious chest injuries for belted occupants involved in rollover crashes requires an understanding of the cause of these injuries and of the test conditions to assure the effectiveness of the countermeasures. This study defines rollover environments and occupant-to-vehicle interactions that cause chest injuries for belted drivers. The NASS-CDS was examined to determine the frequency and crash severity for belted drivers with serious (Abbreviated Injury Scale [AIS] 3+) chest injuries in rollovers. Case studies of NASS crashes with serious chest injuries sustained by belted front occupants were undertaken and damage patterns were determined. Vehicle rollover tests with dummies were examined to determine occupant motion in crashes with damage similar to that observed in the NASS cases. Computer simulations were performed to further explore factors that could contribute to chest injury. Finite element model (FEM) vehicle models with both the FEM Hybrid III dummy and THUMS human model were used in the simulations. Simulation of rollovers with 6 quarter-turns or less indicated that increases in the vehicle pitch, either positive or negative, increased the severity of dummy chest loadings. This finding was consistent with vehicle damage observations from NASS cases. For the far-side occupant, the maximum chest loadings were caused by belt and side interactions during the third quarter-turn and by the center console loading during the fourth quarter-turn. The results showed that the THUMS dummy produced more realistic kinematics and improved insights into skeletal and chest organ loadings compared to the Hybrid III dummy. These results suggest that a dynamic rollover test to encourage chest injury reduction countermeasures should induce a roll of at least 4 quarter-turns and should also include initial vehicle pitch and/or yaw so that the vehicle's axis of rotation is not aligned with its inertial roll axis during the initial stage of the rollover.

75 FR 5931 - Anthropomorphic Test Devices; Hybrid III Test Dummy, ES-2re Side Impact Crash Test Dummy

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-05

... [Docket No. NHTSA-2009-0194] RIN 2127-AK64 Anthropomorphic Test Devices; Hybrid III Test Dummy, ES-2re Side Impact Crash Test Dummy AGENCY: National Highway Traffic Safety Administration (NHTSA), Department... adopted specifications and qualification requirements for a new crash test dummy called the ``ES- 2re...

Orion Crew Member Injury Predictions during Land and Water Landings

NASA Technical Reports Server (NTRS)

Lawrence, Charles; Littell, Justin D.; Fasanella, Edwin L.; Tabiei, Ala

2008-01-01

A review of astronaut whole body impact tolerance is discussed for land or water landings of the next generation manned space capsule named Orion. LS-DYNA simulations of Orion capsule landings are performed to produce a low, moderate, and high probability of injury. The paper evaluates finite element (FE) seat and occupant simulations for assessing injury risk for the Orion crew and compares these simulations to whole body injury models commonly referred to as the Brinkley criteria. The FE seat and crash dummy models allow for varying the occupant restraint systems, cushion materials, side constraints, flailing of limbs, and detailed seat/occupant interactions to minimize landing injuries to the crew. The FE crash test dummies used in conjunction with the Brinkley criteria provides a useful set of tools for predicting potential crew injuries during vehicle landings.

SEMICONDUCTOR TECHNOLOGY Dummy fill effect on CMP planarity

NASA Astrophysics Data System (ADS)

Junxiong, Zhou; Lan, Chen; Wenbiao, Ruan; Zhigang, Li; Weixiang, Shen; Tianchun, Ye

2010-10-01

With the use of a chemical-mechanical polishing (CMP) simulator verified by testing data from a foundry, the effect of dummy fill characteristics, such as fill size, fill density and fill shape, on CMP planarity is analyzed. The results indicate that dummy density has a significant impact on oxide erosion, and copper dishing is in proportion to dummy size. We also demonstrate that cross shape dummy fill can have the best dishing performance at the same density.

CMOS chip planarization by chemical mechanical polishing for a vertically stacked metal MEMS integration

NASA Astrophysics Data System (ADS)

Lee, Hocheol; Miller, Michele H.; Bifano, Thomas G.

2004-01-01

In this paper we present the planarization process of a CMOS chip for the integration of a microelectromechanical systems (MEMS) metal mirror array. The CMOS chip, which comes from a commercial foundry, has a bumpy passivation layer due to an underlying aluminum interconnect pattern (1.8 µm high), which is used for addressing individual micromirror array elements. To overcome the tendency for tilt error in the CMOS chip planarization, the approach is to sputter a thick layer of silicon nitride at low temperature and to surround the CMOS chip with dummy silicon pieces that define a polishing plane. The dummy pieces are first lapped down to the height of the CMOS chip, and then all pieces are polished. This process produced a chip surface with a root-mean-square flatness error of less than 100 nm, including tilt and curvature errors.

Determination of crash test pulses and their application to aircraft seat analysis

NASA Technical Reports Server (NTRS)

Alfaro-Bou, E.; Williams, M. S.; Fasanella, E. L.

1981-01-01

Deceleration time histories (crash pulses) from a series of twelve light aircraft crash tests conducted at NASA Langley Research Center (LaRC) were analyzed to provide data for seat and airframe design for crashworthiness. Two vertical drop tests at 12.8 m/s (42 ft/s) and 36 G peak deceleration (simulating one of the vertical light aircraft crash pulses) were made using an energy absorbing light aircraft seat prototype. Vertical pelvis acceleration measured in a 50 percentile dummy in the energy absorbing seat were found to be 45% lower than those obtained from the same dummy in a typical light aircraft seat. A hybrid mathematical seat-occupant model was developed using the DYCAST nonlinear finite element computer code and was used to analyze a vertical drop test of the energy absorbing seat. Seat and occupant accelerations predicted by the DYCAST model compared quite favorably with experimental values.

ACCELEROMETERS IN FLOW FIELDS: A STRUCTURAL ANALYSIS OF THE CHOPPED DUMMY INPILE TUBE

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

Howard, T. K.; Marcum, W. R.; Latimer, G. D.

2016-06-01

Four tests characterizing the structural response of the Chopped-Dummy In-Pile tube (CDIPT) experiment design were measured in the Hydro-Mechanical Fuel Test Facility (HMFTF). Four different test configurations were tried. These configurations tested the pressure drop and flow impact of various plate configurations and flow control orifices to be used later at different reactor power levels. Accelerometers were placed on the test vehicle and flow simulation housing. A total of five accelerometers were used with one on the top and bottom of the flow simulator and vehicle, and one on the outside of the flow simulator. Data were collected at amore » series of flow rates for 5 seconds each at an acquisition rate of 2 kHz for a Nyquist frequency of 1 kHz. The data were then analyzed using a Fast Fourier Transform (FFT) algorithm. The results show very coherent vibrations of the CDIPT experiment on the order of 50 Hz in frequency and 0.01 m/s2 in magnitude. The coherent vibrations, although small in magnitude pose a potential design problem if the frequencies coincide with the natural frequency of the fueled plates or test vehicle. The accelerometer data was integrated and combined to create a 3D trace of the experiment during the test. The merits of this data as well as further anomalies and artifacts are also discussed as well as their relation to the instrumentation and experiment design.« less

A survey of the dummy face and human face stimuli used in BCI paradigm.

PubMed

Chen, Long; Jin, Jing; Zhang, Yu; Wang, Xingyu; Cichocki, Andrzej

2015-01-15

It was proved that the human face stimulus were superior to the flash only stimulus in BCI system. However, human face stimulus may lead to copyright infringement problems and was hard to be edited according to the requirement of the BCI study. Recently, it was reported that facial expression changes could be done by changing a curve in a dummy face which could obtain good performance when it was applied to visual-based P300 BCI systems. In this paper, four different paradigms were presented, which were called dummy face pattern, human face pattern, inverted dummy face pattern and inverted human face pattern, to evaluate the performance of the dummy faces stimuli compared with the human faces stimuli. The key point that determined the value of dummy faces in BCI systems were whether dummy faces stimuli could obtain as good performance as human faces stimuli. Online and offline results of four different paradigms would have been obtained and comparatively analyzed. Online and offline results showed that there was no significant difference among dummy faces and human faces in ERPs, classification accuracy and information transfer rate when they were applied in BCI systems. Dummy faces stimuli could evoke large ERPs and obtain as high classification accuracy and information transfer rate as the human faces stimuli. Since dummy faces were easy to be edited and had no copyright infringement problems, it would be a good choice for optimizing the stimuli of BCI systems. Copyright © 2014 Elsevier B.V. All rights reserved.

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)

Effect of bottles, cups, and dummies on breast feeding in preterm infants: a randomised controlled trial

PubMed Central

Collins, Carmel T; Ryan, Philip; Crowther, Caroline A; McPhee, Andrew J; Paterson, Susan; Hiller, Janet E

2004-01-01

Objective To determine the effect of artificial teats (bottle and dummy) and cups on breast feeding in preterm infants. Design Randomised controlled trial. Setting Two large tertiary hospitals, 54 peripheral hospitals. Participants 319 preterm infants (born at 23-33 weeks' gestation) randomly assigned to one of four groups: cup/no dummy (n = 89), cup/dummy (n = 72), bottle/no dummy (n = 73), bottle/dummy (n = 85). Women with singleton or twin infants < 34 weeks' gestation who wanted to breastfeed were eligible to participate. Interventions Cup or bottle feeding occurred when the mother was unable to be present to breast feed. Infants randomised to the dummy groups received a dummy on entry into the trial. Main outcome measures Full breast feeding (compared with partial and none) and any breast feeding (compared with none) on discharge home. Secondary outcomes: prevalence of breast feeding at three and six months after discharge and length of hospital stay. Results 303 infants (and 278 mothers) were included in the intention to treat analysis. There were no significant differences for any of the study outcomes according to use of a dummy. Infants randomised to cup feeds were more likely to be fully breast fed on discharge home (odds ratio 1.73, 95% confidence interval 1.04 to 2.88, P = 0.03), but had a longer length of stay (hazard ratio 0.71, 0.55 to 0.92, P = 0.01). Conclusions Dummies do not affect breast feeding in preterm infants. Cup feeding significantly increases the likelihood that the baby will be fully breast fed at discharge home, but has no effect on any breast feeding and increases the length of hospital stay. PMID:15208209

Effect of bottles, cups, and dummies on breast feeding in preterm infants: a randomised controlled trial.

PubMed

Collins, Carmel T; Ryan, Philip; Crowther, Caroline A; McPhee, Andrew J; Paterson, Susan; Hiller, Janet E

2004-07-24

To determine the effect of artificial teats (bottle and dummy) and cups on breast feeding in preterm infants. Randomised controlled trial. Two large tertiary hospitals, 54 peripheral hospitals. 319 preterm infants (born at 23-33 weeks' gestation) randomly assigned to one of four groups: cup/no dummy (n = 89), cup/dummy (n = 72), bottle/no dummy (n = 73), bottle/dummy (n = 85). Women with singleton or twin infants < 34 weeks' gestation who wanted to breastfeed were eligible to participate. Cup or bottle feeding occurred when the mother was unable to be present to breast feed. Infants randomised to the dummy groups received a dummy on entry into the trial. Full breast feeding (compared with partial and none) and any breast feeding (compared with none) on discharge home. prevalence of breast feeding at three and six months after discharge and length of hospital stay. 303 infants (and 278 mothers) were included in the intention to treat analysis. There were no significant differences for any of the study outcomes according to use of a dummy. Infants randomised to cup feeds were more likely to be fully breast fed on discharge home (odds ratio 1.73, 95% confidence interval 1.04 to 2.88, P = 0.03), but had a longer length of stay (hazard ratio 0.71, 0.55 to 0.92, P = 0.01). Dummies do not affect breast feeding in preterm infants. Cup feeding significantly increases the likelihood that the baby will be fully breast fed at discharge home, but has no effect on any breast feeding and increases the length of hospital stay.

Monitoring arrangement for vented nuclear fuel elements

DOEpatents

Campana, Robert J.

1981-01-01

In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

Radial current high power dummy load for characterizing the high power laser triggered transformer-type accelerator.

PubMed

Yin, Yi; Zhong, Hui-Huang; Liu, Jin-Liang; Ren, He-Ming; Yang, Jian-Hua; Zhang, Xiao-Ping; Hong, Zhi-qiang

2010-09-01

A radial-current aqueous resistive solution load was applied to characterize a laser triggered transformer-type accelerator. The current direction in the dummy load is radial and is different from the traditional load in the axial. Therefore, this type of dummy load has smaller inductance and fast response characteristic. The load was designed to accommodate both the resistance requirement of accelerator and to allow optical access for the laser. Theoretical and numerical calculations of the load's inductance and capacitance are given. The equivalent circuit of the dummy load is calculated in theory and analyzed with a PSPICE code. The simulation results agree well with the theoretical analysis. At last, experiments of the dummy load applied to the high power spiral pulse forming line were performed; a quasisquare pulse voltage is obtained at the dummy load.

Radial current high power dummy load for characterizing the high power laser triggered transformer-type accelerator

NASA Astrophysics Data System (ADS)

Yin, Yi; Zhong, Hui-Huang; Liu, Jin-Liang; Ren, He-Ming; Yang, Jian-Hua; Zhang, Xiao-Ping; Hong, Zhi-qiang

2010-09-01

A radial-current aqueous resistive solution load was applied to characterize a laser triggered transformer-type accelerator. The current direction in the dummy load is radial and is different from the traditional load in the axial. Therefore, this type of dummy load has smaller inductance and fast response characteristic. The load was designed to accommodate both the resistance requirement of accelerator and to allow optical access for the laser. Theoretical and numerical calculations of the load's inductance and capacitance are given. The equivalent circuit of the dummy load is calculated in theory and analyzed with a PSPICE code. The simulation results agree well with the theoretical analysis. At last, experiments of the dummy load applied to the high power spiral pulse forming line were performed; a quasisquare pulse voltage is obtained at the dummy load.

Development and validation of a modified Hybrid-III six-year-old dummy model for simulating submarining in motor-vehicle crashes.

PubMed

Hu, Jingwen; Klinich, Kathleen D; Reed, Matthew P; Kokkolaras, Michael; Rupp, Jonathan D

2012-06-01

In motor-vehicle crashes, young school-aged children restrained by vehicle seat belt systems often suffer from abdominal injuries due to submarining. However, the current anthropomorphic test device, so-called "crash dummy", is not adequate for proper simulation of submarining. In this study, a modified Hybrid-III six-year-old dummy model capable of simulating and predicting submarining was developed using MADYMO (TNO Automotive Safety Solutions). The model incorporated improved pelvis and abdomen geometry and properties previously tested in a modified physical dummy. The model was calibrated and validated against four sled tests under two test conditions with and without submarining using a multi-objective optimization method. A sensitivity analysis using this validated child dummy model showed that dummy knee excursion, torso rotation angle, and the difference between head and knee excursions were good predictors for submarining status. It was also shown that restraint system design variables, such as lap belt angle, D-ring height, and seat coefficient of friction (COF), may have opposite effects on head and abdomen injury risks; therefore child dummies and dummy models capable of simulating submarining are crucial for future restraint system design optimization for young school-aged children. Copyright © 2011 IPEM. Published by Elsevier Ltd. All rights reserved.

The effects of dummy/pacifier use on infant blood pressure and autonomic activity during sleep.

PubMed

Yiallourou, Stephanie R; Poole, Hannah; Prathivadi, Pallavi; Odoi, Alexsandria; Wong, Flora Y; Horne, Rosemary S C

2014-12-01

Dummy/pacifier use is protective for sudden infant death syndrome (SIDS); however, the mechanism/s for this are unknown. As impaired cardiovascular control may be the underlying cause of SIDS, we assessed the effects of dummy/pacifier use on cardiovascular control during sleep within the first 6 months of life. Term infants, divided into dummy/pacifier users and non-dummy/pacifier users, were studied at 2-4 weeks (n = 27), 2-3 months (n = 35) and 5-6 months (n = 31) using daytime polysomnography. Heart rate, blood pressure (BP), heart rate variability (HRV), blood pressure variability (BPV), and baroreflex sensitivity (BRS) were measured in triplicate 1-2-min epochs during quiet and active sleep in the supine and prone positions. Overall, during the non-sucking periods, in the prone position, the BP was higher (10-22 mmHg) in dummy/pacifier users compared to non-users at 2-4 weeks and 5-6 months (p < 0.05 for both). HRV and BRS were higher in dummy/pacifier users compared to non-users at 2-4 weeks (p < 0.05). Active sucking increased HRV and BPV, consistent with increased sympathetic activity in dummy/pacifier users. Higher BP and HRV in dummy/pacifier users indicate increased sympathetic tone, which may serve as a protective mechanism against possible hypotension leading to SIDS; however, these effects were not apparent at 2-3 months, when the risk of SIDS is highest. Copyright © 2014 Elsevier B.V. All rights reserved.

Efficient vibration mode analysis of aircraft with multiple external store configurations

NASA Technical Reports Server (NTRS)

Karpel, M.

1988-01-01

A coupling method for efficient vibration mode analysis of aircraft with multiple external store configurations is presented. A set of low-frequency vibration modes, including rigid-body modes, represent the aircraft. Each external store is represented by its vibration modes with clamped boundary conditions, and by its rigid-body inertial properties. The aircraft modes are obtained from a finite-element model loaded by dummy rigid external stores with fictitious masses. The coupling procedure unloads the dummy stores and loads the actual stores instead. The analytical development is presented, the effects of the fictitious mass magnitudes are discussed, and a numerical example is given for a combat aircraft with external wing stores. Comparison with vibration modes obtained by a direct (full-size) eigensolution shows very accurate coupling results. Once the aircraft and stores data bases are constructed, the computer time for analyzing any external store configuration is two to three orders of magnitude less than that of a direct solution.

Method of determining whether radioactive contaminants are inside or outside a structure

DOEpatents

Lattin, Kenneth R.

1977-01-01

A measure is obtained of the relative quantities of radioactive material inside and outside a structure such as a pipe by obtaining two spectra of gamma radiation on a dummy structure of the same shape and composition. A first spectrum is obtained with a quantity of the radioactive element to be measured located inside the structure and a second spectrum is obtained with a quantity of the same contaminant located outside the structure. The two spectra are normalized to the same equivalent value in a portion of the spectrum that does not reflect the presence of gamma rays resulting from Compton scattering in the structure. Comparison of that portion of the spectra obtained where Compton scattering is a factor gives a measure of the relative amounts of contaminants inside and outside the structure on a spectrum obtained from a test structure. The invention may also be practiced by obtaining a plurality of spectra at varying known concentrations inside and outside the dummy structure.

FUEL ELEMENT SUPPORT

DOEpatents

Wyman, W.L.

1961-06-27

The described cylindrical fuel element has longitudinally spaced sets of short longitudinal ribs circumferentially spaced from one another. The ribs support the fuel element in a coolant tube so that there is an annular space for coolant flow between the fuel element and the interior of the coolant tube. If the fuel element grows as a result of reactor operation, the circumferential distribution of the ribs maintains the uniformity of the annular space between the coolant tube and the fuel element, and the collapsibility of the ribs prevents the fuel element from becoming jammed in the coolant tube.

Thermal breeder fuel enrichment zoning

DOEpatents

Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

1992-01-01

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

75 FR 71648 - Federal Motor Vehicle Safety Standards, Child Restraint Systems; Hybrid III 10-Year-Old Child...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-24

...This document proposes to amend Federal Motor Vehicle Safety Standard (FMVSS) No. 213, Child Restraint Systems, regarding a Hybrid III 10-year-old child test dummy that the agency seeks to use in the compliance test procedures of the standard. This document supplements a 2005 notice of proposed rulemaking (NPRM) and a 2008 SNPRM previously published in this rulemaking (RIN 2127-AJ44) regarding this test dummy. In the 2005 NPRM, in response to Anton's Law, NHTSA proposed to adopt the 10-year-old child test dummy into FMVSS No. 213 to test child restraints for older children. Subsequently, to address variation that was found in dummy readings due to chin-to-chest contact, NHTSA published the 2008 SNPRM to propose a NHTSA-developed procedure for positioning the test dummy in belt-positioning seats. Comments on the SNPRM objected to the positioning procedure, and some suggested an alternative procedure developed by the University of Michigan Transportation Research Institute (UMTRI). Today's SNPRM proposes to use the UMTRI procedure to position the test dummy rather than the NHTSA-developed procedure. We note that the 10-year-old child dummy may sometimes experience stiff contact between its chin and upper sternal bib region which may result in an unrealistically high value of the head injury criterion (HIC) \\1\\ referenced in the standard. Accordingly, NHTSA proposes that the dummy's HIC measurement will not be used to assess the compliance of the tested child restraint. This SNPRM also proposes other amendments to FMVSS No. 213, including a proposal to permit NHTSA to use, at the manufacturer's option, the Hybrid II or Hybrid III versions of the 6-year-old test dummy, and a proposal to use the UMTRI procedure to position the Hybrid III 6-year- old and 10-year-old dummies when testing belt-positioning seats. ---------------------------------------------------------------------------

A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

Fire Protection of Weapon Storage and Water Mist Redundancy Philosophies

DTIC Science & Technology

2012-11-01

criteria me system ged system ozzles dummy tor d, insulated titute of Swe stems pedo pipe Date 2012 den Refere -03-31 P90 nce 0038-04...test wit tion test wit ution test wi t system, 10 st system, 5 m, 5 bar, 50 , 10 bar, 50 ummy, free- edo dummy pedo dummy pedo dummy ummy, dren...systems usi lower volum pedo dumm temperature discharge d ion. h Institute ynamics dström Date 2012 den ater mist/wa ests indicate fire

Low temperature chemical processing of graphite-clad nuclear fuels

DOEpatents

Pierce, Robert A.

2017-10-17

A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

Dummy left behind by Skylab 3 crew for the Skylab 4 crew

NASA Technical Reports Server (NTRS)

1973-01-01

This photograph is an illustration of the humorous side of the Skylab 3 crew. This dummy was left behind in the Skylab space station by the Skylab 3 crew to be found by the Skylab 4 crew. The dummy is dressed in a flight suit and placed in the Lower Body Negative Pressure Device. The name tag indicates that it represents Gerald P. Carr, Skylab 4 commander. In the background is a partial view of the dummy for William R. Pogue, Skylab 4 pilot, propped upon the bicycle ergometer (1586); This dummy is dressed in a flight suit and propped upon the bicycle ergometer. The name tag indicates that it represents William R. Pogue, Skylab 4 pilot (1587).

77 FR 5418 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-03

... aft fuel system 40 micron fuel filter element with a 10 micron fuel filter element. This proposed AD... fuel filter element, part number (P/N) 52-0505-2 or 52-01064-1. This proposed AD would require replacing each forward and aft fuel system 40 micron fuel filter element with a 10 micron fuel filter...

Tetramorium tsushimae Ants Use Methyl Branched Hydrocarbons of Aphids for Partner Recognition.

PubMed

Sakata, Itaru; Hayashi, Masayuki; Nakamuta, Kiyoshi

2017-10-01

In mutualisms, partner discrimination is often the most important challenge for interacting organisms. The interaction between ants and aphids is a model system for studying mutualisms; ants are provided with honeydew by aphids and, in turn, the ants offer beneficial services to the aphids. To establish and maintain this system, ants must discriminate mutualistic aphid species correctly. Although recent studies have shown that ants recognize aphids as mutualistic partners based on their cuticular hydrocarbons (CHCs), it was unclear which CHCs are involved in recognition. Here, we tested whether the n-alkane or methylalkane fraction, or both, of aphid CHCs were utilized as partner recognition cues by measuring ant aggressiveness toward these fractions. When workers of Tetramorium tsushimae ants were presented with dummies coated with n-alkanes of their mutualistic aphid Aphis craccivora, ants displayed higher levels of aggression than to dummies treated with total CHCs or methyl alkanes of A. craccivora; responses to dummies treated with n-alkanes of A. craccivora were similar to those to control dummies or dummies treated with the CHCs of the non-mutualistic aphid Acyrthosiphon pisum. By contrast, ants exhibited lower aggression to dummies treated with either total CHCs or the methylalkane fraction of the mutualistic aphid than to control dummies or dummies treated with CHCs of the non-mutualistic aphid. These results suggest that T. tsushimae ants use methylalkanes of the mutualistic aphid's CHCs to recognize partners, and that these ants do not recognize aphids as partners on the basis of n-alkanes.

49 CFR 572.197 - Abdomen.

Code of Federal Regulations, 2012 CFR

2012-10-01

..., DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES IIsD Side Impact Crash Test Dummy... impacted side removed. The dummy is equipped with a lower spine laterally oriented accelerometer as... side of the seated dummy tangent to a vertical plane located within 10 mm of the side edge of the bench...

49 CFR 572.197 - Abdomen.

Code of Federal Regulations, 2011 CFR

2011-10-01

..., DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES IIsD Side Impact Crash Test Dummy... impacted side removed. The dummy is equipped with a lower spine laterally oriented accelerometer as... side of the seated dummy tangent to a vertical plane located within 10 mm of the side edge of the bench...

NUCLEAR REACTOR FUEL ELEMENT

DOEpatents

Wheelock, C.W.; Baumeister, E.B.

1961-09-01

A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

49 CFR 572.191 - General description.

Code of Federal Regulations, 2011 CFR

2011-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES IIsD Side Impact Crash Test Dummy, Small Adult Female § 572.191 General description. (a) The SID-IIsD Side Impact Crash Test Dummy... the SID-IIsD Side Impact Crash Test Dummy, 5th percentile adult female, is shown in drawing 180-0000...

49 CFR 572.191 - General description.

Code of Federal Regulations, 2012 CFR

2012-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES IIsD Side Impact Crash Test Dummy, Small Adult Female § 572.191 General description. (a) The SID-IIsD Side Impact Crash Test Dummy... the SID-IIsD Side Impact Crash Test Dummy, 5th percentile adult female, is shown in drawing 180-0000...

49 CFR 572.197 - Abdomen.

Code of Federal Regulations, 2013 CFR

2013-10-01

..., DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES SID-IIsD Side Impact Crash Test Dummy... impacted side removed. The dummy is equipped with a lower spine laterally oriented accelerometer as... side of the seated dummy tangent to a vertical plane located within 10 mm of the side edge of the bench...

49 CFR 572.199 - Pelvis iliac.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 7 2013-10-01 2013-10-01 false Pelvis iliac. 572.199 Section 572.199... Test Dummy, Small Adult Female § 572.199 Pelvis iliac. (a) The iliac is part of the lower torso... the assembled dummy (drawing 180-0000). The dummy is equipped with a laterally oriented pelvis...

49 CFR 572.199 - Pelvis iliac.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 7 2011-10-01 2011-10-01 false Pelvis iliac. 572.199 Section 572.199... Dummy, Small Adult Female § 572.199 Pelvis iliac. (a) The iliac is part of the lower torso assembly... assembled dummy (drawing 180-0000). The dummy is equipped with a laterally oriented pelvis accelerometer as...

49 CFR 572.199 - Pelvis iliac.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 7 2014-10-01 2014-10-01 false Pelvis iliac. 572.199 Section 572.199... Test Dummy, Small Adult Female § 572.199 Pelvis iliac. (a) The iliac is part of the lower torso... the assembled dummy (drawing 180-0000). The dummy is equipped with a laterally oriented pelvis...

49 CFR 572.198 - Pelvis acetabulum.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 7 2012-10-01 2012-10-01 false Pelvis acetabulum. 572.198 Section 572.198... Dummy, Small Adult Female § 572.198 Pelvis acetabulum. (a) The acetabulum is part of the lower torso... torso of the assembled dummy (drawing 180-0000). The dummy is equipped with a laterally oriented pelvis...

49 CFR 572.199 - Pelvis iliac.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 7 2010-10-01 2010-10-01 false Pelvis iliac. 572.199 Section 572.199... Dummy, Small Adult Female § 572.199 Pelvis iliac. (a) The iliac is part of the lower torso assembly... assembled dummy (drawing 180-0000). The dummy is equipped with a laterally oriented pelvis accelerometer as...

49 CFR 572.199 - Pelvis iliac.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 7 2012-10-01 2012-10-01 false Pelvis iliac. 572.199 Section 572.199... Dummy, Small Adult Female § 572.199 Pelvis iliac. (a) The iliac is part of the lower torso assembly... assembled dummy (drawing 180-0000). The dummy is equipped with a laterally oriented pelvis accelerometer as...

49 CFR 572.198 - Pelvis acetabulum.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 7 2011-10-01 2011-10-01 false Pelvis acetabulum. 572.198 Section 572.198... Dummy, Small Adult Female § 572.198 Pelvis acetabulum. (a) The acetabulum is part of the lower torso... torso of the assembled dummy (drawing 180-0000). The dummy is equipped with a laterally oriented pelvis...

49 CFR 572.161 - General description.

Code of Federal Regulations, 2010 CFR

2010-10-01

... Weighted Child Test Dummy § 572.161 General description. (a) The Hybrid III Six-Year-Old Weighted Child Test Dummy is defined by drawings and specifications containing the following materials: (1) “Parts List and Drawings, Part 572 Subpart S, Hybrid III Weighted Six-Year Old Child Test Dummy (H-III6CW...

Dummy left behind by Skylab 3 crew for the Skylab 4 crew

NASA Image and Video Library

1973-08-16

SL3-113-1586 (July-September 1973) --- This photograph is an illustration of the humorous side of the Skylab 3 crew. This dummy was left behind in the Skylab space station by the Skylab 3 crew to be found by the Skylab 4 crew. The dummy is dressed in a flight suit and placed in the Lower Body Negative Pressure Device. The name tag indicates that it represents Gerald P. Carr, Skylab 4 commander, in the background is a partial view of the dummy for William R. Pogue, Skylab 4 pilot, propped upon the bicycle ergometer. The dummy representing Edward G. Gibson, Skylab science pilot, was left in the waste compartment. Astronauts Alan L. Bean, Owen K. Garriott and Jack R. Lousma were the Skylab 3 crewmen. Photo credit: NASA

Dummy left behind by Skylab 3 crew for the Skylab 4 crew

NASA Image and Video Library

1973-08-16

SL3-113-1587 (July-September 1973) --- This photograph is an illustration of the humorous side of the Skylab 3 crew. This dummy was left behind in the Skylab space station by the Skylab 3 crew to be found by the Skylab 4 crew. The dummy is dressed in a flight suit and propped upon the bicycle ergometer. The name tag indicated that it represents William R. Pogue, Skylab pilot. The dummy for Gerald P. Carr, Skylab 4 commander, was placed in the Lower Body Negative Pressure Device. The dummy representing Edward G. Gibson was left in the waste compartment. Astronauts Alan L. Bean, Owen K. Garriott and Jack R. Lousma were the Skylab 3 crewmen. Gibson is the Skylab 4 science pilot. Photo credit: NASA

Male and female WorldSID and post mortem human subject responses in full-scale vehicle tests.

PubMed

Yoganandan, Narayan; Humm, John; Pintar, Frank; Rhule, Heather; Moorhouse, Kevin; Suntay, Brian; Stricklin, Jim; Rudd, Rodney; Craig, Matthew

2017-05-29

This study compares the responses of male and female WorldSID dummies with post mortem human subject (PMHS) responses in full-scale vehicle tests. Tests were conducted according to the FMVSS-214 protocols and using the U.S. Side Impact New Car Assessment Program change in velocity to match PMHS experiments, published earlier. Moving deformable barrier (MDB) tests were conducted with the male and female surrogates in the left front and left rear seats. Pole tests were performed with the male surrogate in the left front seat. Three-point belt restraints were used. Sedan-type vehicles were used from the same manufacturer with side airbags. The PMHS head was instrumented with a pyramid-shaped nine-axis accelerometer package, with angular velocity transducers on the head. Accelerometers and angular velocity transducers were secured to T1, T6, and T12 spinous processes and sacrum. Three chest bands were secured around the upper, middle, and lower thoraces. Dummy instrumentation included five infrared telescoping rods for assessment of chest compression (IR-TRACC) and a chest band at the first abdomen rib, head angular velocity transducer, and head, T1, T4, T12, and pelvis accelerometers. Morphological responses of the kinematics of the head, thoracic spine, and pelvis matched in both surrogates for each pair. The peak magnitudes of the torso accelerations were lower for the dummy than for the biological surrogate. The brain rotational injury criterion (BrIC) response was the highest in the male dummy for the MDB test and PMHS. The probability of AIS3+ injuries, based on the head injury criterion, ranged from 3% to 13% for the PMHS and from 3% to 21% for the dummy from all tests. The BrIC-based metrics ranged from 0 to 21% for the biological and 0 to 48% for the dummy surrogates. The deflection profiles from the IR-TRACC sensors were unimodal. The maximum deflections from the chest band placed on the first abdominal rib were 31.7 mm and 25.4 mm for the male and female dummies in the MDB test, and 37.4 mm for the male dummy in the pole test. The maximum deflections computed from the chest band contours at a gauge equivalent to the IR-TRACC location were 25.9 mm and 14.8 mm for the male and female dummies in the MDB test, and 37.4 mm for the male dummy in the pole test. Other data (static vehicle deformation profiles, accelerations histories of different body regions, and chest band contours for the dummy and PMHS) are given in the appendix. This is the first study to compare the responses of PMHS and male and female dummies in MDB and pole tests, done using the same recent model year vehicles with side airbag and head curtain restraints. The differences between the dummy and PMHS torso accelerations suggest the need for design improvements in the WorldSID dummy. The translation-based metrics suggest low probability of head injury. As the dummy internal sensor underrecorded the peak deflection, multipoint displacement measures are therefore needed for a more accurate quantification of deflection to improve the safety assessment of occupants.

Low cost, lightweight fuel cell elements

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor)

2001-01-01

New fuel cell elements for use in liquid feed fuel cells are provided. The elements including biplates and endplates are low in cost, light in weight, and allow high efficiency operation. Electrically conductive elements are also a part of the fuel cell elements.

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1958-08-19

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

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.

Injury risk curves for the WorldSID 50th male dummy.

PubMed

Petitjean, Audrey; Trosseille, Xavier; Petit, Philippe; Irwin, Annette; Hassan, Joe; Praxl, Norbert

2009-11-01

The development of the WorldSID 50th percentile male dummy was initiated in 1997 by the International Organisation for Standardisation (ISO/SC12/TC22/WG5) with the objective of developing a more biofidelic side impact dummy and supporting the adoption of a harmonised dummy into regulations. More than 45 organizations from all around the world have contributed to this effort including governmental agencies, research institutes, car manufacturers and dummy manufacturers. The first production version of the WorldSID 50th male dummy was released in March 2004 and demonstrated an improved biofidelity over existing side impact dummies. Full scale vehicle tests covering a wide range of side impact test procedures were performed worldwide with the WorldSID dummy. However, the vehicle safety performance could not be assessed due to lack of injury risk curves for this dummy. The development of these curves was initiated in 2004 within the framework of ISO/SC12/TC22/WG6 (Injury criteria). In 2008, the ACEA- Dummy Task Force (TFD) decided to contribute to this work and offered resources for a project manager to coordinate of the effort of a group of volunteer biomechanical experts from international institutions (ISO, EEVC, VRTC/NHTSA, JARI, Transport Canada), car manufacturers (ACEA, Ford, General Motors, Honda, Toyota, Chrysler) and universities (Wayne State University, Ohio State University, John Hopkins University, Medical College of Wisconsin) to develop harmonized injury risk curves. An in-depth literature review was conducted. All the available PMHS datasets were identified, the test configurations and the quality of the results were checked. Criteria were developed for inclusion or exclusion of PMHS tests in the development of the injury risk curves. Data were processed to account for differences in mass and age of the subjects. Finally, injury risk curves were developed using the following statistical techniques, the certainty method, the Mertz/Weber method, the logistic regression, the survival analysis and the Consistent Threshold Estimate. The paper presents the methods used to check and process the data, select the PMHS tests, and construct the injury risk curves. The PMHS dataset as well as the injury risk curves are provided.

Method of locating a leaking fuel element in a fast breeder power reactor

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

49 CFR 572.198 - Pelvis acetabulum.

Code of Federal Regulations, 2010 CFR

2010-10-01

... the dummy is in vertical orientation. (4) Push the dummy at the knees and at mid-sternum of the upper torso with just sufficient horizontally oriented force towards the seat back until the back of the upper torso is in contact with the seat back. (5) While maintaining the dummy's position as specified in...

49 CFR 572.195 - Thorax with arm.

Code of Federal Regulations, 2010 CFR

2010-10-01

... dummy is in vertical orientation. (4) Push the dummy at the knees and at mid-sternum of the upper torso with just sufficient horizontally oriented force towards the seat back until the back of the upper torso is in contact with the seat back. (5) While maintaining the dummy's position as specified in...

How Robust Is Linear Regression with Dummy Variables?

ERIC Educational Resources Information Center

Blankmeyer, Eric

2006-01-01

Researchers in education and the social sciences make extensive use of linear regression models in which the dependent variable is continuous-valued while the explanatory variables are a combination of continuous-valued regressors and dummy variables. The dummies partition the sample into groups, some of which may contain only a few observations.…

Dummy Cup Helps Robot-Welder Programmers

NASA Technical Reports Server (NTRS)

Gordon, Stephen S.

1990-01-01

Dummy gas cup used on torch of robotic welder during programming and practice runs. Made of metal or plastic, dummy cup inexpensive and durable. Withstands bumps caused by programming errors, and is sized for special welding jobs within limited clearances. After robot satisfactorily programmed, replaced by ceramic cup of same dimensions for actual welding.

CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

DOEpatents

Weems, S.J.

1963-09-24

A rib construction for a nuclear-fuel element is described, in which one of three peripherally spaced ribs adjacent to each end of the fuel element is mounted on a radially yielding spring that embraces the fuel element. This spring enables the fuel element to have a good fit with a coolant tube and yet to be easily inserted in and withdrawn from the tube. (AEC)

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

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

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

49 CFR 572.140 - Incorporation by reference.

Code of Federal Regulations, 2012 CFR

2012-10-01

... Child Crash Test Dummy, Alpha Version § 572.140 Incorporation by reference. (a) The following materials... entitled, “Parts List and Drawings, Subpart P Hybrid III 3-year-old child crash test dummy, (H-III3C, Alpha..., Disassembly and Inspection (PADI), Subpart P, Hybird III 3-year-old Child Crash Test Dummy, (H-III3C, Alpha...

49 CFR 572.140 - Incorporation by reference.

Code of Federal Regulations, 2011 CFR

2011-10-01

... Child Crash Test Dummy, Alpha Version § 572.140 Incorporation by reference. (a) The following materials... entitled, “Parts List and Drawings, Subpart P Hybrid III 3-year-old child crash test dummy, (H-III3C, Alpha..., Disassembly and Inspection (PADI), Subpart P, Hybird III 3-year-old Child Crash Test Dummy, (H-III3C, Alpha...

49 CFR 572.140 - Incorporation by reference.

Code of Federal Regulations, 2014 CFR

2014-10-01

... Child Crash Test Dummy, Alpha Version § 572.140 Incorporation by reference. (a) The following materials... entitled, “Parts List and Drawings, Subpart P Hybrid III 3-year-old child crash test dummy, (H-III3C, Alpha..., Disassembly and Inspection (PADI), Subpart P, Hybird III 3-year-old Child Crash Test Dummy, (H-III3C, Alpha...

49 CFR 572.140 - Incorporation by reference.

Code of Federal Regulations, 2013 CFR

2013-10-01

... Child Crash Test Dummy, Alpha Version § 572.140 Incorporation by reference. (a) The following materials... entitled, “Parts List and Drawings, Subpart P Hybrid III 3-year-old child crash test dummy, (H-III3C, Alpha..., Disassembly and Inspection (PADI), Subpart P, Hybird III 3-year-old Child Crash Test Dummy, (H-III3C, Alpha...

49 CFR 572.140 - Incorporation by reference.

Code of Federal Regulations, 2010 CFR

2010-10-01

... Child Crash Test Dummy, Alpha Version § 572.140 Incorporation by reference. (a) The following materials... entitled, “Parts List and Drawings, Subpart P Hybrid III 3-year-old child crash test dummy, (H-III3C, Alpha..., Disassembly and Inspection (PADI), Subpart P, Hybird III 3-year-old Child Crash Test Dummy, (H-III3C, Alpha...

49 CFR 572.76 - Limbs assembly and test procedure.

Code of Federal Regulations, 2010 CFR

2010-10-01

... between 1g and 2g. (ii) Place the dummy legs in a plane parallel to the dummy's midsagittal plane with the knee pivot center line perpendicular to the dummy's midsagittal plane, and with the feet flat on the... parallel to the midsagittal plane at the specified velocity. (5) Guide the test probe during impact so that...

TLINES: A Computer Program for Circuits of Transmission Lines.

DTIC Science & Technology

1983-12-01

of various lengths are handled by stringing together many short lines, with the assumption that each of the longer lines has a length approximated as...expressed in terms of transmission lines numbered from 2 through CAPM , connected in numerical sequence as in figure 3. Line 1 is a dummy element disconnected...from line 2 and the rest of the circuit. Lines 2 through CAPM can each be set to any impedance the user desires. Line CAPM +1 is a zero-impedance line

Laboratory Reconstructions of Real World Frontal Crash Configurations using the Hybrid III and THOR Dummies and PMHS.

PubMed

Petitjean, Audrey; Lebarbe, Matthieu; Potier, Pascal; Trosseille, Xavier; Lassau, Jean-Pierre

2002-11-01

Load-limiting belt restraints have been present in French cars since 1995. An accident study showed the greater effectiveness in thorax injury prevention using a 4 kN load limiter belt with an airbag than using a 6 kN load limiter belt without airbag. The criteria for thoracic tolerance used in regulatory testing is the sternal deflection for all restraint types, belt and/or airbag restraint. This criterion does not assess the effectiveness of the restraint 4 kN load limiter belt with airbag observed in accidentology. To improve the understanding of thoracic tolerance, frontal sled crashes were performed using the Hybrid III and THOR dummies and PMHS. The sled configuration and the deceleration law correspond to those observed in the accident study. Restraint conditions evaluated are the 6 kN load-limiting belt and the 4 kN load-limiting belt with an airbag. Loads between the occupant and the sled environment were recorded. Various measurements (including thoracic deflections and head, thorax and pelvis accelerations and angular velocities on the dummies) characterize the dummy and PMHS behavior. PMHS anthropometry and injuries were noted. This study presents the test methodology and the results used to evaluate dummy ability to discriminate both restraint types and dummy measurement ability to be representative of thoracic injury risk for all restraint types. The injury results of the PMHS tests showed the same tendency as the accident study. Some of the criteria proposed in the literature did not show a better protection of the 4 kN load limiter belt with airbag restraint, in particular thoracic deflection maxima for both dummies. The four thoracic deflections measured on the THOR and Hybrid III dummies may allow more accurate analysis of the loading pattern and therefore of injury risk.

Trauma potential and ballistic parameters of cal. 9 mm P.A. dummy launchers.

PubMed

Frank, Matthias; Bockholdt, Britta; Philipp, Klaus-Peter; Ekkernkamp, Axel

2010-07-15

Blank cartridge actuated dummy launching devices are used by migratory bird hunters to train dogs to retrieve downed birds. The devices create a loud noise while simultaneously propelling a hard foam dummy for retrieval. A newly developed dummy launcher is based on a modified cal. 9 mm P.A. blank handgun with an extension tube pinned and welded to the barrel imitation. Currently, there are no experimental investigations on the ballistic background and trauma potential of these uncommon shooting devices. An experimental test set-up consisting of a photoelectric infrared light barrier was used for measurement of the velocity of hard foam dummies propelled with an automatic dummy launcher. Ballistic parameters of the dummies and an aluminium sleeve as improvised projectile (kinetic energy (E), impulse (p), energy density (E') and threshold velocity (v(tsh)) to cause penetrating wounds as a function of cross-sectional density (S)) were calculated. The average velocity (v) of the dummies was measured 25.71 m/s exerting an average impulse (p) of 3.342 Ns. The average kinetic energy (E) was calculated 43.04 J with an average energy density (E') of 0.069 J/mm(2). The average velocity (v) of the aluminium sleeves as improvised projectiles was measured 79.58 m/s exerting an average impulse (p) of 2.228 Ns. The average kinetic energy (E) of the aluminium sleeves was calculated as 88.70 J with an average energy density (E') of 0.282 J/mm(2). The energy delivered by these shooting devices is high enough to cause relevant injuries. The absence of skin penetration must not mislead the emergency physician or forensic expert into neglecting the potential damage from these devices. (c) 2010 Elsevier Ireland Ltd. All rights reserved.

Intrathoracic pressure variations in an anthropomorphic dummy exposed to air blast, blunt impact, and missiles.

PubMed

Jönsson, A; Arvebo, E; Schantz, B

1988-01-01

Experiments with an anthropomorphic dummy for blast research demonstrated that pressures recorded in the lung model of the dummy could be correlated to primary air blast effects on the lungs of experimental animals. The results presented here were obtained with a dummy of the type mentioned above, but with the lung model modified to improve geometric similarity to man. Blast experiments were performed in a shock tube, and impact experiments in a special impact machine. Experiments with nonpenetrating missiles were performed with small-caliber firearms and the dummy protected by body armor. Severity indices derived from the blast experiments were related to established criteria for primary lung injury in man. Impacts delivered in the impact machine and by nonpenetrating missiles are compared. Relationships between severity of impact based on experiments with animals and primary lung injury in man are discussed.

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1962-08-14

A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

Transition Core Properties during Conversion of the NBSR from HEU to LEU Fuel

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

Hanson, A. L.; Diamond, D.

2013-10-31

The transition of the NBSR from HEU to LEU fuel is challenging due to reactivity constraints and the need to maintain an uninterrupted science program, the mission of the NBSR. The transition cannot occur with a full change of HEU to LEU fuel elements since the excess reactivity would be large enough that the NBSR would violate the technical specification for shutdown margin. Manufacturing LEU fuel elements to represent irradiated fuel elements would be cost prohibitive since 26 one-of-a-kind fuel elements would need to be manufactured. For this report a gradual transition from the present HEU fuel to the proposedmore » LEU fuel was studied. The gradual change approach would follow the present fuel management scheme and replace four HEU fuel elements with four LEU fuel elements each cycle. This manuscript reports the results of a series of calculations to predict the neutronic characteristics and how the neutronics will change during the transition from HEU to LEU in the NBSR.« less

Numerical reconstruction and injury biomechanism in a car-pedestrian crash accident.

PubMed

Zou, Dong-Hua; Li, Zheng-Dong; Shao, Yu; Feng, Hao; Chen, Jian-Guo; Liu, Ning-Guo; Huang, Ping; Chen, Yi-Jiu

2012-12-01

To reconstruct a car-pedestrian crash accident using numerical simulation technology and explore the injury biomechanism as forensic evidence for injury identification. An integration of multi-body dynamic, finite element (FE), and classical method was applied to a car-pedestrian crash accident. The location of the collision and the details of the traffic accident were determined by vehicle trace verification and autopsy. The accident reconstruction was performed by coupling the three-dimensional car behavior from PC-CRASH with a MADYMO dummy model. The collision FE models of head and leg, developed from CT scans of human remains, were loaded with calculated dummy collision parameters. The data of the impact biomechanical responses were extracted in terms of von Mises stress, relative displacement, strain and stress fringes. The accident reconstruction results were identical with the examined ones and the biomechanism of head and leg injuries, illustrated through the FE methods, were consistent with the classical injury theories. The numerical simulation technology is proved to be effective in identifying traffic accidents and exploring of injury biomechanism.

75 FR 76636 - Anthropomorphic Test Devices; Hybrid III 6-Year-Old Child Test Dummy, Hybrid III 6-Year-Old...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-09

... test H06120 with the original femurs. Therefore, comparisons were made between pre- and post-test... [Docket No. NHTSA-2010-0147] RIN 2127-AK34 Anthropomorphic Test Devices; Hybrid III 6-Year-Old Child Test Dummy, Hybrid III 6-Year-Old Weighted Child Test Dummy AGENCY: National Highway Traffic Safety...

NEUTRONIC REACTOR

DOEpatents

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Nuclear reactor

DOEpatents

Thomson, Wallace B.

2004-03-16

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

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.

Nuclear reactor control

DOEpatents

Cawley, William E.; Warnick, Robert F.

1982-01-01

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Dummy Measurement of Chest Injuries Induced by Two-Point Shoulder Belts

PubMed Central

Augenstein, J.; Perdeck, E.; Bowen, J.; Stratton, J.; Horton, T.; Singer, M.; Digges, K.; Malliaris, A.; Steps, J.

2000-01-01

The University of Miami’s William Lehman Injury Research Center at the Jackson Memorial Medical Center conducts interdisciplinary investigations to study seriously injured restrained occupants in frontal automobile collisions. Engineering analysis of these crashes is conducted in conjunction with the National Crash Analysis Center at the George Washington University. The multidisciplinary research team includes expertise in crash investigation, crash reconstruction, computer graphics, biomechanics of injuries, crash data analysis, trauma care, and all of the medical specialties associated with the Ryder Trauma Center at Jackson Memorial Hospital. More than 350 injured occupants and their crashes have been studied in depth. The purpose of this paper is to report on an observed pattern of liver lacerations suffered by drivers wearing shoulder belts, without the lap belt fastened and to assess the ability of existing crash test dummies to measure the potential for these injuries. During the initial years of the study, 48 cases of drivers protected by shoulder belts but without the lap belt fastened met the criteria for the study. Fifty percent of these drivers suffered liver lacerations. Further study showed that 22 of the crashes involved damage to the right front of the vehicle. Among the drivers in vehicles with right front damage, 92% sustained injuries to the liver. This observation indicated that 2-point belts were most likely to produce liver injuries in low severity frontal collisions when the crash direction is 1 to 2 o’clock. An analysis of the National Accident Sampling System for the years 1988-95 indicated that liver injuries constitute about 0.5% of the injuries suffered by drivers who are in tow-away crashes. NASS data showed that the risk of chest injury is more likely among drivers with automatic shoulder belts than drivers with 3-point manual belts. The crash test dummies showed no difference in chest injury measures. Finite element computer modeling demonstrated that the high deflection of the right lower rib on the Hybrid III dummy predicts the liver injuries in the 1 o’clock crashes. These higher deflections were less apparent at the location of the center chest deflection measurement device on the Hybrid III. PMID:11558077

Dummy measurement of chest injuries induced by two-point shoulder belts.

PubMed

Augenstein, J; Perdeck, E; Bowen, J; Stratton, J; Horton, T; Singer, M; Digges, K; Malliaris, A; Steps, J

2000-01-01

The University of Miami's William Lehman Injury Research Center at the Jackson Memorial Medical Center conducts interdisciplinary investigations to study seriously injured restrained occupants in frontal automobile collisions. Engineering analysis of these crashes is conducted in conjunction with the National Crash Analysis Center at the George Washington University. The multidisciplinary research team includes expertise in crash investigation, crash reconstruction, computer graphics, biomechanics of injuries, crash data analysis, trauma care, and all of the medical specialties associated with the Ryder Trauma Center at Jackson Memorial Hospital. More than 350 injured occupants and their crashes have been studied in depth. The purpose of this paper is to report on an observed pattern of liver lacerations suffered by drivers wearing shoulder belts, without the lap belt fastened and to assess the ability of existing crash test dummies to measure the potential for these injuries. During the initial years of the study, 48 cases of drivers protected by shoulder belts but without the lap belt fastened met the criteria for the study. Fifty percent of these drivers suffered liver lacerations. Further study showed that 22 of the crashes involved damage to the right front of the vehicle. Among the drivers in vehicles with right front damage, 92% sustained injuries to the liver. This observation indicated that 2-point belts were most likely to produce liver injuries in low severity frontal collisions when the crash direction is 1 to 2 o'clock. An analysis of the National Accident Sampling System for the years 1988-95 indicated that liver injuries constitute about 0.5% of the injuries suffered by drivers who are in tow-away crashes. NASS data showed that the risk of chest injury is more likely among drivers with automatic shoulder belts than drivers with 3-point manual belts. The crash test dummies showed no difference in chest injury measures. Finite element computer modeling demonstrated that the high deflection of the right lower rib on the Hybrid III dummy predicts the liver injuries in the 1 o'clock crashes. These higher deflections were less apparent at the location of the center chest deflection measurement device on the Hybrid III.

77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

Impact Testing and Simulation of a Crashworthy Composite Fuselage Section with Energy-Absorbing Seats and Dummies

NASA Technical Reports Server (NTRS)

Fasanella, Edwin L.; Jackson, Karen E.

2002-01-01

A 25-ft/s vertical drop test of a composite fuselage section was conducted with two energy-absorbing seats occupied by anthropomorphic dummies to evaluate the crashworthy features of the fuselage section and to determine its interaction with the seats and dummies. The 5-ft. diameter fuselage section consists of a stiff structural floor and an energy-absorbing subfloor constructed of Rohacel foam blocks. The experimental data from this test were analyzed and correlated with predictions from a crash simulation developed using the nonlinear, explicit transient dynamic computer code, MSC.Dytran. The anthropomorphic dummies were simulated using the Articulated Total Body (ATB) code, which is integrated into MSC.Dytran.

Impact Testing and Simulation of a Crashworthy Composite Fuselage Section with Energy-Absorbing Seats and Dummies

NASA Technical Reports Server (NTRS)

Fasanella, Edwin L.; Jackson, Karen E.

2002-01-01

A 25-ft/s vertical drop test of a composite fuselage section was conducted with two energy-absorbing seats occupied by anthropomorphic dummies to evaluate the crashworthy features of the fuselage section and to determine its interaction with the seats and dummies. The 5-ft diameter fuselage section consists of a stiff structural floor and an energy-absorbing subfloor constructed of Rohacel foam blocks. The experimental data from this test were analyzed and correlated with predictions from a crash simulation developed using the nonlinear, explicit transient dynamic computer code, MSC.Dytran. The anthropomorphic dummies were simulated using the Articulated Total Body (ATB) code, which is integrated into MSC.Dytran.

In-situ membrane hydration measurement of proton exchange membrane fuel cells

NASA Astrophysics Data System (ADS)

Lai, Yeh-Hung; Fly, Gerald W.; Clapham, Shawn

2015-01-01

Achieving proper membrane hydration control is one of the most critical aspects of PEM fuel cell development. This article describes the development and application of a novel 50 cm2 fuel cell device to study the in-situ membrane hydration by measuring the through-thickness membrane swelling via an array of linear variable differential transducers. Using this setup either as an air/air (dummy) cell or as a hydrogen/air (operating) cell, we performed a series of hydration and dehydration experiments by cycling the RH of the inlet gas streams at 80 °C. From the linear relationship between the under-the-land swelling and the over-the-channel water content, the mechanical constraint within the fuel cell assembly can suppress the membrane water uptake by 11%-18%. The results from the air/air humidity cycling test show that the membrane can equilibrate within 120 s for all RH conditions and that membrane can reach full hydration at a RH higher than 140% in spite of the use of a liquid water impermeable Carbel MP30Z microporous layer. This result confirms that the U.S. DOE's humidity cycling mechanical durability protocol induces sufficient humidity swings to maximize hygrothermal mechanical stresses. This study shows that the novel experimental technique can provide a robust and accurate means to study the in-situ hydration of thin membranes subject to a wide range of fuel cell conditions.

Anthropometry for WorldSID, a World-Harmonized Midsize Male Side Impact Crash Dummy

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

S. Moss; Z. Wang; M. Salloum

2000-06-19

The WorldSID project is a global effort to design a new generation side impact crash test dummy under the direction of the International Organization for Standardization (ISO). The first WorldSID crash dummy will represent a world-harmonized mid-size adult male. This paper discusses the research and rationale undertaken to define the anthropometry of a world standard midsize male in the typical automotive seated posture. Various anthropometry databases are compared region by region and in terms of the key dimensions needed for crash dummy design. The Anthropometry for Motor Vehicle Occupants (AMVO) dataset, as established by the University of Michigan Transportation Researchmore » Institute (UMTRI), is selected as the basis for the WorldSID mid-size male, updated to include revisions to the pelvis bone location. The proposed mass of the dummy is 77.3kg with full arms. The rationale for the selected mass is discussed. The joint location and surface landmark database is appended to this paper.« less

Fuel pumping system and method

DOEpatents

Shafer, Scott F [Morton, IL; Wang, Lifeng ,

2006-12-19

A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

Fuel Pumping System And Method

DOEpatents

Shafer, Scott F.; Wang, Lifeng

2005-12-13

A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

Object categorization by wild ranging birds-Winter feeder experiments.

PubMed

Nováková, Nela; Veselý, Petr; Fuchs, Roman

2017-10-01

The object categorization is only scarcely studied using untrained wild ranging animals and relevant stimuli. We tested the importance of the spatial position of features salient for categorization of a predator using wild ranging birds (titmice) visiting a winter feeder. As a relevant stimulus we used a dummy of a raptor, the European sparrowhawk (Accipiter nisus), placed at the feeding location. This dummy was designed to be dismantled into three parts and rearranged with the head in the correct position, in the middle or at the bottom of the dummy. When the birds had the option of visiting an alternative feeder with a dummy pigeon, they preferred this option to visiting the feeder with the dummy sparrowhawk with the head in any of the three positions. When the birds had the option of visiting an alternative feeder with an un-rearranged dummy sparrowhawk, they visited both feeders equally often, and very scarcely. This suggests that the titmice considered all of the sparrowhawk modifications as being dangerous, and equally dangerous as the un-rearranged sparrowhawk. The position of the head was not the most important cue for categorization. The presence of the key features was probably sufficient for categorization, and their mutual spatial position was of lower importance. Copyright © 2017 Elsevier B.V. All rights reserved.

Preparation of "dummy" l-phenylalanine molecularly imprinted microspheres by using ionic liquid as a template and functional monomer.

PubMed

Li, Ji; Hu, Xiaoling; Guan, Ping; Song, Dongmen; Qian, Liwei; Du, Chunbao; Song, Renyuan; Wang, Chaoli

2015-07-07

In this study, dummy imprinting technology was employed for the preparation of l-phenylalanine-imprinted microspheres. Ionic liquids were utilized as both a "dummy" template and functional monomer, and 4-vinylpyridine and ethylene glycol dimethacrylate were used as the assistant monomer and cross-linker, respectively, for preparing a surface-imprinted polymer on poly(divinylbenzene) microspheres. By the results obtained by theoretical investigation, the interaction between the template and monomer complex was improved as compared with that between the template and the traditional l-phenylalanine-imprinted polymer. The batch experiments indicated that the imprinting factor reached 2.5. Scatchard analysis demonstrated that the obtained "dummy" molecularly imprinted microspheres exhibited an affinity of 77.4 M·10 -4 , significantly higher that of a traditional polymer directly prepared by l-phenylalanine, which is in agreement with theoretical results. Competitive adsorption experiments also showed that the molecularly imprinted polymer with the dummy template effectively isolated l-phenylalanine from l-histidine and l-tryptophan with separation factors of 5.68 and 2.68, respectively. All these results demonstrated that the polymerizable ionic liquid as the dummy template could enhance the affinity and selectivity of molecularly imprinted polymer, thereby promoting the development of imprinting technology for biomolecules. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Means for supporting fuel elements in a nuclear reactor

DOEpatents

Andrews, Harry N.; Keller, Herbert W.

1980-01-01

A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

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.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

NASA Technical Reports Server (NTRS)

Emrich, William

2009-01-01

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

Current status of the development of high density LEU fuel for Russian research reactors

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

Vatulin, A.; Dobrikova, I.; Suprun, V.

2008-07-15

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element

DTIC Science & Technology

1990-06-01

long fuel elements, arranged to form a core , were analyzed for an up-power transient from 0 MWt to approximately 18 MWt. The simple model significantly...VARIATIONS IN FUEL ELEMENT GEOMETRY ............. 60 4.4 VARIATIONS IN THE MANNER OF TRANSIENT CONTROL ..... 62 4.5 CORE REPRESENTATION BY MULTIPLE FUEL ...the HTGR , however, the PBR packs small fuel particles between inner and outer retention elements, designated as frits. The PBR is appropriate for a

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

78 FR 17591 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-22

... aft fuel system 40 micron fuel filter element with a 10 micron nominal (40 micron absolute) fuel filter element. This AD was prompted by a National Transportation Safety Board (NTSB) review of in... helicopters with a fuel system 40 micron fuel filter element, part number (P/N) 52-0505-2 or 52-01064-1. That...

Cleanup Verification Package for the 118-F-1 Burial Ground

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

E. J. Farris and H. M. Sulloway

2008-01-10

This cleanup verification package documents completion of remedial action for the 118-F-1 Burial Ground on the Hanford Site. This burial ground is a combination of two locations formerly called Minor Construction Burial Ground No. 2 and Solid Waste Burial Ground No. 2. This waste site received radioactive equipment and other miscellaneous waste from 105-F Reactor operations, including dummy elements and irradiated process tubing; gun barrel tips, steel sleeves, and metal chips removed from the reactor; filter boxes containing reactor graphite chips; and miscellaneous construction solid waste.

NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

DOEpatents

Szilard, L.; Young, G.J.

1958-03-01

This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

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

Montierth, Leland M.

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

The repeatability and reproducibility of the BioRID IIg in a repeatable laboratory seat based on a production car seat.

PubMed

Hynd, David; Depinet, Paul; Lorenz, Bernd

2013-01-01

The United Nations Economic Commission for Europe Informal Group on GTR No. 7 Phase 2 are working to define a build level for the BioRID II rear impact (whiplash) crash test dummy that ensures repeatable and reproducible performance in a test procedure that has been proposed for future legislation. This includes the specification of dummy hardware, as well as the development of comprehensive certification procedures for the dummy. This study evaluated whether the dummy build level and certification procedures deliver the desired level of repeatability and reproducibility. A custom-designed laboratory seat was made using the seat base, back, and head restraint from a production car seat to ensure a representative interface with the dummy. The seat back was reinforced for use in multiple tests and the recliner mechanism was replaced by an external spring-damper mechanism. A total of 65 tests were performed with 6 BioRID IIg dummies using the draft GTR No.7 sled pulse and seating procedure. All dummies were subject to the build, maintenance, and certification procedures defined by the Informal Group. The test condition was highly repeatable, with a very repeatable pulse, a well-controlled seat back response, and minimal observed degradation of seat foams. The results showed qualitatively reasonable repeatability and reproducibility for the upper torso and head accelerations, as well as for T1 Fx and upper neck Fx . However, reproducibility was not acceptable for T1 and upper neck Fz or for T1 and upper neck My . The Informal Group has not selected injury or seat assessment criteria for use with BioRID II, so it is not known whether these channels would be used in the regulation. However, the ramping-up behavior of the dummy showed poor reproducibility, which would be expected to affect the reproducibility of dummy measurements in general. Pelvis and spine characteristics were found to significantly influence the dummy measurements for which poor reproducibility was observed. It was also observed that the primary neck response in these tests was flexion, not extension. This correlates well with recent findings from Japan and the United States showing a correlation between neck flexion and injury in accident replication simulations and postmortem human subjects (PMHS) studies, respectively. The present certification tests may not adequately control front cervical spine bumper characteristics, which are important for neck flexion response. The certification sled test also does not include the pelvis and so cannot be used to control pelvis response and does not substantially load the lumbar bumpers and so does not control these parts of the dummy. The stiffness of all spine bumpers and of the pelvis flesh should be much more tightly controlled. It is recommended that a method for certifying the front cervical bumpers should be developed. Recommendations are also made for tighter tolerance on the input parameters for the existing certification tests.

Parameter screening: the use of a dummy parameter to identify non-influential parameters in a global sensitivity analysis

NASA Astrophysics Data System (ADS)

Khorashadi Zadeh, Farkhondeh; Nossent, Jiri; van Griensven, Ann; Bauwens, Willy

2017-04-01

Parameter estimation is a major concern in hydrological modeling, which may limit the use of complex simulators with a large number of parameters. To support the selection of parameters to include in or exclude from the calibration process, Global Sensitivity Analysis (GSA) is widely applied in modeling practices. Based on the results of GSA, the influential and the non-influential parameters are identified (i.e. parameters screening). Nevertheless, the choice of the screening threshold below which parameters are considered non-influential is a critical issue, which has recently received more attention in GSA literature. In theory, the sensitivity index of a non-influential parameter has a value of zero. However, since numerical approximations, rather than analytical solutions, are utilized in GSA methods to calculate the sensitivity indices, small but non-zero indices may be obtained for the indices of non-influential parameters. In order to assess the threshold that identifies non-influential parameters in GSA methods, we propose to calculate the sensitivity index of a "dummy parameter". This dummy parameter has no influence on the model output, but will have a non-zero sensitivity index, representing the error due to the numerical approximation. Hence, the parameters whose indices are above the sensitivity index of the dummy parameter can be classified as influential, whereas the parameters whose indices are below this index are within the range of the numerical error and should be considered as non-influential. To demonstrated the effectiveness of the proposed "dummy parameter approach", 26 parameters of a Soil and Water Assessment Tool (SWAT) model are selected to be analyzed and screened, using the variance-based Sobol' and moment-independent PAWN methods. The sensitivity index of the dummy parameter is calculated from sampled data, without changing the model equations. Moreover, the calculation does not even require additional model evaluations for the Sobol' method. A formal statistical test validates these parameter screening results. Based on the dummy parameter screening, 11 model parameters are identified as influential. Therefore, it can be denoted that the "dummy parameter approach" can facilitate the parameter screening process and provide guidance for GSA users to define a screening-threshold, with only limited additional resources. Key words: Parameter screening, Global sensitivity analysis, Dummy parameter, Variance-based method, Moment-independent method

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

Full-Scale Transport controlled Impact Demonstration

NASA Technical Reports Server (NTRS)

Hayduk, R. J. (Compiler)

1986-01-01

The controlled impact demonstration (CID) test of a transport aircraft took place on December 1, 1984, crashing at a prepared site on Rogers Dry Lakebed, Edwards Air Force Base, California. The demonstration was a setback for the antimisting kerosene (AMK) researchers. The impact conditions, considerably different from the planned scenario, exposed large quantities of degraded AMK and hydraulic fluid and caused unexpectedly hot ignition sources, bulk loss of fuel from the right wing, airflow patterns over the wings and fuselage that were untested on AMK, and fuel intrusion into the lower fuselage. The test was much more severe than planned and is generally considered to be unrepresentative of the type of survivable crash that would benefit from AMK. Ninety-seven percent of the sensors on the fuselage and wing structure, seats, dummies, restraint systems, galley, and bins were active at impact. A wealth of sensor data was collected from this once-in-a-lifetime research test. The flight data recorder experiments on board were also generally successful.

Evaluation of the finite element fuel rod analysis code (FRANCO)

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

Lee, K.; Feltus, M.A.

1994-12-31

Knowledge of temperature distribution in a nuclear fuel rod is required to predict the behavior of fuel elements during operating conditions. The thermal and mechanical properties and performance characteristics are strongly dependent on the temperature, which can vary greatly inside the fuel rod. A detailed model of fuel rod behavior can be described by various numerical methods, including the finite element approach. The finite element method has been successfully used in many engineering applications, including nuclear piping and reactor component analysis. However, fuel pin analysis has traditionally been carried out with finite difference codes, with the exception of Electric Powermore » Research Institute`s FREY code, which was developed for mainframe execution. This report describes FRANCO, a finite element fuel rod analysis code capable of computing temperature disrtibution and mechanical deformation of a single light water reactor fuel rod.« less

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.

New type of dummy layout pattern to control ILD etch rate

NASA Astrophysics Data System (ADS)

Pohland, Oliver; Spieker, Julie; Huang, Chih-Ta; Govindaswamy, Srikanth; Balasinski, Artur

2007-12-01

Adding dummy features (waffles) to drawn geometries of the circuit layout is a common practice to improve its manufacturability. As an example, local dummy pattern improves MOSFET line and space CD control by adjusting short range optical proximity and reducing the aggressiveness of its correction features (OPC) to widen the lithography process window. Another application of dummy pattern (waffles) is to globally equalize layout pattern density, to reduce long-range inter-layer dielectric (ILD) thickness variations after the CMP process and improve contact resistance uniformity over the die area. In this work, we discuss a novel type of dummy pattern with a mid-range interaction distance, to control the ILD composition driven by its deposition and etch process. This composition is reflected on sidewall spacers and depends on the topography of the underlying poly pattern. During contact etch, it impacts the etch rate of the ILD. As a result, the deposited W filling the damascene etched self-aligned trench contacts in the ILD may electrically short to the underlying gates in the areas of isolated poly. To mitigate the dependence of the ILD composition on poly pattern distribution, we proposed a special dummy feature generation with the interaction range defined by the ILD deposition and etch process. This helped equalize mid-range poly pattern density without disabling the routing capability with damascene trench contacts in the periphery which would have increased the layout footprint.

Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer Coefficient (HTC), bulk-fluid, sheath and fuel centerline temperature profiles were generated along the heated length of 5.772 m for a generic fuel channel. The fuel centerline and sheath temperature profiles have been determined at four Axial Heat Flux Profiles (AHFPs) using an average thermal power per channel of 8.5 MWth. The four examined AHFPs are the uniform, cosine, upstream-skewed and downstream-skewed profiles. Additionally, this study focuses on investigating a possibility of using low, enhanced and high thermal-conductivity fuels. The low thermal-conductivity fuels, which have been examined in this study, are uranium dioxide (UO 2), Mixed Oxide (MOX) and Thoria (ThO2) fuels. The examined enhanced thermal-conductivity fuels are uranium dioxide - silicon carbide (UO2 - SiC) and uranium dioxide - beryllium oxide (UO2 - BeO). Lastly, uranium carbide (UC), uranium dicarbide (UC2) and uranium nitride (UN) are the selected high thermal-conductivity fuels, which have been proposed for use in SCWRs. A comparison has been made between the low, enhanced and high thermal-conductivity fuels in order to identify the fuel centerline temperature behaviour when different nuclear fuels are used. Also, in the process of conducting the sensitivity analysis, the HTC was calculated using the Mokry et al. correlation, which is the most accurate supercritical water heat-transfer correlation so far. The sheath and the fuel centerline temperature profiles were determined for two cases. In Case 1, the HTC was calculated based on the Mokry et al. correlation, while in Case 2, the HTC values calculated for Case 1 were multiplied by a factor of 2. This factor was used in order to identify the amount of decrease in temperatures if the heat transfer is enhanced with appendages. Results of this analysis indicate that the use of the newly developed 54-element fuel bundle along with the proposed fuels is promising when compared with the Variant-20 (43-element) fuel bundle. Overall, the fuel centerline and sheath temperatures were below the industry and design limits when most of the proposed fuels were examined in the 54-element fuel bundle, however, the fuel centerline temperature limit was exceeded while MOX fuel was examined. Keywords: SCWRs, Fuel Centerline Temperature, Sheath Temperature, High Thermal Conductivity Fuels, Low Thermal Conductivity Fuels, HTC.

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

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

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

Development of a shear force measurement dummy for seat comfort.

PubMed

Kim, Seong Guk; Ko, Chang-Yong; Kim, Dong Hyun; Song, Ye Eun; Kang, Tae Uk; Ahn, Sungwoo; Lim, Dohyung; Kim, Han Sung

2017-01-01

Seat comfort is one of the main factors that consumers consider when purchasing a car. In this study, we develop a dummy with a shear-force sensor to evaluate seat comfort. The sensor has dimensions of 25 mm × 25 mm × 26 mm and is made of S45C. Electroless nickel plating is employed to coat its surface in order to prevent corrosion and oxidation. The proposed sensor is validated using a qualified load cell and shows high accuracy and precision (measurement range: -30-30 N; sensitivity: 0.1 N; linear relationship: R = 0.999; transverse sensitivity: <1%). The dummy is manufactured in compliance with the SAE standards (SAE J826) and incorporates shear sensors into its design. We measure the shear force under four driving conditions and at five different speeds using a sedan; results showed that the shear force increases with speed under all driving conditions. In the case of acceleration and deceleration, shear force significantly changes in the lower body of the dummy. During right and left turns, it significantly changes in the upper body of the dummy.

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)

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY

DOEpatents

Wigner, E.P.; Young, G.J.; Weinberg, A.M.

1961-06-27

A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-16

A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

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.

SLUG HANDLING DEVICES

DOEpatents

Gentry, J.R.

1958-09-16

A device is described for handling fuel elements of a neutronic reactor. The device consists of two concentric telescoped contalners that may fit about the fuel element. A number of ratchet members, equally spaced about the entrance to the containers, are pivoted on the inner container and spring biased to the outer container so thnt they are forced to hear against and hold the fuel element, the weight of which tends to force the ratchets tighter against the fuel element. The ratchets are released from their hold by raising the inner container relative to the outer memeber. This device reduces the radiation hazard to the personnel handling the fuel elements.

FUEL ELEMENTS FOR NUCLEAR REACTORS

DOEpatents

Blainey, A.; Lloyd, H.

1961-07-11

A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

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

Trammell, Michael P; Jolly, Brian C; Miller, James Henry

ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

Fuel element concept for long life high power nuclear reactors

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

Full-Scale Crash Test of a MD-500 Helicopter with Deployable Energy Absorbers

NASA Technical Reports Server (NTRS)

Kellas, Sotiris; Jackson, Karen E.; Littell, Justin D.

2010-01-01

A new externally deployable energy absorbing system was demonstrated during a full-scale crash test of an MD-500 helicopter. The deployable system is a honeycomb structure and utilizes composite materials in its construction. A set of two Deployable Energy Absorbers (DEAs) were fitted on the MD-500 helicopter for the full-scale crash demonstration. Four anthropomorphic dummy occupants were also used to assess human survivability. A demonstration test was performed at NASA Langley's Landing and Impact Research Facility (LandIR). The test involved impacting the helicopter on a concrete surface with combined forward and vertical velocity components of 40-ft/s and 26-ft/s, respectively. The objectives of the test were to evaluate the performance of the DEA concept under realistic crash conditions and to generate test data for validation of dynamic finite element simulations. Descriptions of this test as well as other component and full-scale tests leading to the helicopter test are discussed. Acceleration data from the anthropomorphic dummies showed that dynamic loads were successfully attenuated to within non-injurious levels. Moreover, the airframe itself survived the relatively severe impact and was retested to provide baseline data for comparison for cases with and without DEAs.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ...

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

35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ELEMENT HOLDER, TRIP MECHANISM COVER, AND OTHER DETAILS. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-3. INEL INDEX CODE NUMBER: 075 0701 60 851 151977. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Possible consequences of operation with KIVN fuel elements in K Zircaloy process tubes

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

Carlson, P.A.

1963-08-06

From considerations of the results of experimental simulations of non-axial placement of fuel elements in process tubes and in-reactor experience, it is concluded that the ultimate outcome of a charging error which results in operation with one or more unsupported fuel elements in a K Zircaloy-2 process tube would be multiple fuel failure and failure of the process tube. The outcome of the accident is determined by the speed with which the fuel failure is detected and the reactor is shut down. The release of fission products would be expected to be no greater than that which has occurred followingmore » severe fuel failure incidents. The highest probability for fission product release occurs during the discharge of failed fuel elements, when a small fraction of the exposed uranium of the fuel element may be oxidized when exposed to air before the element falls into the water-filled discharge chute. The confinement and fog spray facilities were installed to reduce the amount of fission products which might escape from the reactor building after such an event.« less

Influence of standing or seated pelvis on dummy responses in rear impacts.

PubMed

Viano, David C; Parenteau, Chantal S; Burnett, Roger

2012-03-01

There is a question whether the standing or seated pelvis should be used in Hybrid III dummy evaluations of seats and belt restraint systems in severe rear impacts. This study compares the standing and seated Hybrid III pelvis in matched rear sled tests. Sixteen sled tests were found at 10, 16 and 24 km/h rear delta V in Ford's archives where matched tests were run with the standing and seated pelvis in a belted Hybrid III dummy. Two new tests were conducted at 40 km/h rear delta V to extend the severity range. The head, chest and pelvis were instrumented with triaxial accelerometers and the upper and lower neck, thoracic spine and lumbar spine had transducers measuring triaxial loads and moments. Belt Loads were measured. High-speed video recorded different views of the dummy motion. Dummy kinematics and biomechanical responses were compared for all of the data with the two different Hybrid III pelvic designs. In the 40 km/h sled tests, the dummy motion and excursion were essentially similar with the standing and seated pelvis. The similarities included the lap belt interaction with the pelvis and the leg movement upward flexing the hip joint. Overall, similar biomechanic and kinematic responses were found, including the pelvic acceleration, spinal forces and moments. For the lower speed tests at 10, 16 and 24 km/h, the motion sequence was also similar with the two different pelvises, including the upward movement of the legs as the seat was loaded and rebound kinematics. The biomechanical responses were similar. The seated pelvis involves only a small portion of the upper leg molded into the vinyl skin of the pelvis and does not limit leg rotation at the hip joint. Furthermore, lap belt loads were minimal during the rearward movement of the dummy. The matched testing showed no significant difference in occupant kinematics or biomechanical responses between the standing and seated pelvis in rear sled tests. The Hybrid III dummy with the seated pelvis is suitable for FMVSS 301 and other testing of seats and belt restraint systems in severe rear impacts. Copyright © 2011 Elsevier Ltd. All rights reserved.

Rack for storing spent nuclear fuel elements

DOEpatents

Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

1978-06-20

A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

METHOD OF OPERATING NUCLEAR REACTORS

DOEpatents

Untermyer, S.

1958-10-14

A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

DOEpatents

Sease, J.D.; Harrington, F.E.

1973-12-11

Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

49 CFR 572.41 - General description.

Code of Federal Regulations, 2013 CFR

2013-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES Side Impact Dummy 50th... set forth in the Side Impact Dummy (SID) User's Manual, dated May 1994 except for pages 7, 20 and 23...

49 CFR 572.41 - General description.

Code of Federal Regulations, 2011 CFR

2011-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES Side Impact Dummy 50th... set forth in the Side Impact Dummy (SID) User's Manual, dated May 1994 except for pages 7, 20 and 23...

49 CFR 572.41 - General description.

Code of Federal Regulations, 2010 CFR

2010-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES Side Impact Dummy 50th... set forth in the Side Impact Dummy (SID) User's Manual, dated May 1994 except for pages 7, 20 and 23...

49 CFR 572.41 - General description.

Code of Federal Regulations, 2014 CFR

2014-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES Side Impact Dummy 50th... set forth in the Side Impact Dummy (SID) User's Manual, dated May 1994 except for pages 7, 20 and 23...

49 CFR 572.41 - General description.

Code of Federal Regulations, 2012 CFR

2012-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES Side Impact Dummy 50th... set forth in the Side Impact Dummy (SID) User's Manual, dated May 1994 except for pages 7, 20 and 23...

Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

NASA Technical Reports Server (NTRS)

Mireles, O. R.; Broadway, J.; Hickman, R.

2014-01-01

Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

Average male and female virtual dummy model (BioRID and EvaRID) simulations with two seat concepts in the Euro NCAP low severity rear impact test configuration.

PubMed

Linder, Astrid; Holmqvist, Kristian; Svensson, Mats Y

2018-05-01

Soft tissue neck injuries, also referred to as whiplash injuries, which can lead to long term suffering accounts for more than 60% of the cost of all injuries leading to permanent medical impairment for the insurance companies, with respect to injuries sustained in vehicle crashes. These injuries are sustained in all impact directions, however they are most common in rear impacts. Injury statistics have since the mid-1960s consistently shown that females are subject to a higher risk of sustaining this type of injury than males, on average twice the risk of injury. Furthermore, some recently developed anti-whiplash systems have revealed they provide less protection for females than males. The protection of both males and females should be addresses equally when designing and evaluating vehicle safety systems to ensure maximum safety for everyone. This is currently not the case. The norm for crash test dummies representing humans in crash test laboratories is an average male. The female part of the population is not represented in tests performed by consumer information organisations such as NCAP or in regulatory tests due to the absence of a physical dummy representing an average female. Recently, the world first virtual model of an average female crash test dummy was developed. In this study, simulations were run with both this model and an average male dummy model, seated in a simplified model of a vehicle seat. The results of the simulations were compared to earlier published results from simulations run in the same test set-up with a vehicle concepts seat. The three crash pulse severities of the Euro NCAP low severity rear impact test were applied. The motion of the neck, head and upper torso were analysed in addition to the accelerations and the Neck Injury Criterion (NIC). Furthermore, the response of the virtual models was compared to the response of volunteers as well as the average male model, to that of the response of a physical dummy model. Simulations with the virtual male and female dummy models revealed differences in dynamic response related to the crash severity, as well as between the two dummies in the two different seat models. For the comparison of the response of the virtual models to the response of the volunteers and the physical dummy model, the peak angular motion of the first thoracic vertebra as found in the volunteer tests and mimicked by the physical dummy were not of the same magnitude in the virtual models. The results of the study highlight the need for an extended test matrix that includes an average female dummy model to evaluate the level of occupant protection different seats provide in vehicle crashes. This would provide developers with an additional tool to ensure that both male and female occupants receive satisfactory protection and promote seat concepts that provide the best possible protection for the whole adult population. This study shows that using the mathematical models available today can provide insights suitable for future testing. Copyright © 2017 Elsevier Ltd. All rights reserved.

Injury risk curves for the WorldSID 50th male dummy.

PubMed

Petitjean, Audrey; Trosseille, Xavier; Praxl, Norbert; Hynd, David; Irwin, Annette

2012-10-01

The development of the WorldSID 50th percentile male dummy was initiated in 1997 by the International Organisation for Standardisation (ISO/TC22/SC12/WG5) with the objective of developing a more biofidelic side impact dummy and supporting the adoption of a harmonised dummy into regulations. The dummy is currently under evaluation at the Working Party on Passive Safety (GRSP) in order to be included in the pole side impact global technical regulation (GTR). Injury risk curves dedicated to this dummy and built on behalf of ISO/TC22/SC12/WG6 were proposed in order to assess the occupant safety performance (Petitjean et al. 2009). At that time, there was no recommendation yet on the injury criteria and no consensus on the most accurate statistical method to be used. Since 2009, ISO/TC22/SC12/WG6 reached a consensus on the definition of guidelines to build injury risk curves, including the use of the survival analysis, the distribution assessment and quality checks. These guidelines were applied to the WorldSID 50th results published in 2009 in order to be able to provide a final set of injury risk curves recommended by ISO/TC22/SC12/WG6. The paper presents the different steps of the guidelines as well as the recommended injury risk curves dedicated to the WorldSID 50th for lateral shoulder load, thoracic rib deflection, abdomen rib deflection and pubic force.

Upper and Lower Neck Loads in Belted Human Surrogates in Frontal Impacts

PubMed Central

Yoganandan, Narayan; Pintar, Frank A.; Moore, Jason; Rinaldi, James; Schlick, Michael; Maiman, Dennis J.

2012-01-01

The upper and lower neck loads in the restrained Hybrid III dummy and Test Device for Human Occupant Restraint (THOR) were computed in simulated frontal impact sled tests at low, medium, and high velocities; repeatability performance of the two dummies were evaluated at all energy inputs; peak forces and moments were compared with computed loads at the occipital condyles and cervical-thoracic junctions from tests using post mortem human surrogates (PMHS). A custom sled buck was used to position the surrogates. Repeated tests were conducted at each velocity for each dummy and sufficient time was allowed to elapse between the two experiments. The upper and lower neck forces and moments were determined from load cell measures and its locations with respect to the ends of the neck. Both dummies showed good repeatability for axial and shear forces and bending moments at all changes in velocity inputs. Morphological characteristics in the neck loading responses were similar in all surrogates, although the peak magnitudes of the variables differed. In general, the THOR better mimicked the PMHS response than the Hybrid III dummy, and factors such as neck design and chest compliance were attributed to the observed variations. While both dummies were not designed for use at the two extremes of the tested velocities, results from the present study indicate that, currently the THOR may be the preferred anthropomorphic testing device in crashworthiness research studies and full-scale vehicle tests at all velocities. PMID:23169123

Preparation of high temperature gas-cooled reactor fuel element

DOEpatents

Bradley, Ronnie A.; Sease, John D.

1976-01-01

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

Nuclear fuel element

DOEpatents

Zocher, Roy W.

1991-01-01

A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

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

Konyashov, Vadim V.; Krasnov, Alexander M.

Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. Anmore » approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)« less

49 CFR 572.194 - Shoulder.

Code of Federal Regulations, 2010 CFR

2010-10-01

..., while the midsagittal plane of the dummy is in vertical orientation. (4) Push the dummy at the knees and... back until the back of the upper torso is in contact with the seat back. (5) While maintaining the...

Characterization of deformable materials in the THOR dummy

DOT National Transportation Integrated Search

2000-01-01

Methodologies used to characterize the mechanical behavior of various materials used in the construction of the crash test dummy called THOR (Test device for Human Occupant Restraint) are described. These materials include polyurethane, neoprene, and...

49 CFR 572.91 - General description.

Code of Federal Regulations, 2010 CFR

2010-10-01

... structural properties of the dummy are such that the dummy conforms to this part in every respect both before and after being used in dynamic tests specified in Standard No. 213 of this chapter (§ 571.213). ...

49 CFR 572.91 - General description.

Code of Federal Regulations, 2011 CFR

2011-10-01

... structural properties of the dummy are such that the dummy conforms to this part in every respect both before and after being used in dynamic tests specified in Standard No. 213 of this chapter (§ 571.213). ...

49 CFR 572.81 - General description.

Code of Federal Regulations, 2011 CFR

2011-10-01

... contacts that exist under static conditions. (c) The structural properties of the dummy are such that the dummy conforms to this part in every respect both before and after being used in dynamic tests such as...

Photographic combustion characterization of LOX/Hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1980-01-01

One hundred twenty-seven tests were conducted over a chamber pressure range of 125-1500 psia, a fuel temperature range of -245 F to 158 F, and a fuel velocity range of 48-707 ft/sec to demonstrate the advantages and limitations of using high speed photography to identify potential combustion anomalies such as pops, fuel freezing, reactive stream separation and carbon formations. Combustion evaluation criteria were developed to guide selection of the fuels, injector elements, and operating conditions for testing. Separate criteria were developed for fuel and injector element selection and evaluation. The photographic test results indicated conclusively that injector element type and design directly influence carbon formation. Unlike spray fan, impingement elements reduce carbon formation because they induce a relatively rapid near zone fuel vaporization rate. Coherent jet impingement elements, on the other hand, exhibit increased carbon formation.

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

The application of dummy noise adaptive Kalman filter in underwater navigation

NASA Astrophysics Data System (ADS)

Li, Song; Zhang, Chun-Hua; Luan, Jingde

2011-10-01

The track of underwater target is easy to be affected by the various by the various factors, which will cause poor performance in Kalman filter with the error in the state and measure model. In order to solve the situation, a method is provided with dummy noise compensative technology. Dummy noise is added to state and measure model artificially, and then the question can be solved by the adaptive Kalman filter with unknown time-changed statistical character. The simulation result of underwater navigation proves the algorithm is effective.

NEUTRONIC REACTOR CHARGING AND DISCHARGING

DOEpatents

Zinn, W.H.

1959-07-14

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Fuel handling apparatus for a nuclear reactor

DOEpatents

Hawke, Basil C.

1987-01-01

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Drying results of K-Basin fuel element 1990 (Run 1)

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

Marschman, S.C.; Abrefah, J.; Klinger, G.S.

1998-06-01

The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtainedmore » from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.« less

FUEL ELEMENT FOR NEUTRONIC REACTORS

DOEpatents

Evans, T.C.; Beasley, E.G.

1961-01-17

A fuel element for neutronic reactors, particularly the gas-cooled type of reactor, is described. The element comprises a fuel-bearing plate rolled to form a cylinder having a spiral passageway passing from its periphery to its center. In operation a coolant is admitted to the passageway at the periphery of the element, is passed through the spiral passageway, and emerges into a central channel defined by the inner turn of the rolled plate. The advantage of the element is that the fully heated coolant (i.e., coolant emerging into the central channel) is separated and thus insulated from the periphery of the element, which may be in contact with a low-temperature moderator, by the intermediate turns of the spiral fuel element.

Deflections from two types of Human Surrogates in Oblique Side Impacts

PubMed Central

Yoganandan, Narayan; Pintar, Frank A.

2008-01-01

The objective of the study was to obtain time-dependent thoracic and abdominal deflections of an anthropomorphic test device, the WorldSID dummy, in oblique impact using sled tests, and compare with post mortem human subject (PMHS) data. To simulate the oblique loading vector, the load wall was configured such that the thorax and abdominal plates were offset by twenty or thirty degrees. Deflections were obtained from a chestband placed at the middle thoracic level and five internal deflection transducers. Data were compared from the chestband and the transducer located at the same level of the thorax. In addition, data were compared with deflections from similar PMHS tests obtained using chestbands placed at the level of the axilla, xyphoid process, and tenth rib, representing the upper thorax, middle thorax, and abdominal region of the biological specimen. Peak deflections ranged from 30 to 85 mm in the dummy tests. Peak deflections ranged from 60 to 115 mm in PMHS. Under both obliquities, dummy deflection-time histories at the location along the chestband in close proximity to the internal deflection transducer demonstrated similar profiles. However, the peak deflection magnitudes from the chestband were approximately 20 mm greater than those from the internal transducer. Acknowledging that the chestband measures external deflections in contrast to the transducer which records internal ribcage deformations, peak deflections match from the two sensors. Deflection time histories were also similar between the dummy and PMHS in terms of morphology, although thoracic deflection magnitudes from the dummy matched more closely with PMHS than abdominal deflection magnitudes. The dummy deformed in such a way that peak deflections occurred along the lateral vector. This was in contrast to PMHS tests wherein maximum deflections occurred along the antero-lateral direction, suggesting differing deformation responses in the two models. In addition, peak deflections occurred earlier in the dummy than in PMHS. These preliminary results are valuable in future crashworthiness studies. PMID:19026246

Social interactions between live and artificial weakly electric fish: Electrocommunication and locomotor behavior of Mormyrus rume proboscirostris towards a mobile dummy fish

PubMed Central

Kirschbaum, Frank; von der Emde, Gerhard

2017-01-01

Mormyrid weakly electric fish produce short, pulse-type electric organ discharges for actively probing their environment and to communicate with conspecifics. Animals emit sequences of pulse-trains that vary in overall frequency and temporal patterning and can lead to time-locked interactions with the discharge activity of other individuals. Both active electrolocation and electrocommunication are additionally accompanied by stereotypical locomotor patterns. However, the concrete roles of electrical and locomotor patterns during social interactions in mormyrids are not well understood. Here we used a mobile fish dummy that was emitting different types of electrical playback sequences to study following behavior and interaction patterns (electrical and locomotor) between individuals of weakly electric fish. We confronted single individuals of Mormyrus rume proboscirostris with a mobile dummy fish designed to attract fish from a shelter and recruit them into an open area by emitting electrical playbacks of natural discharge sequences. We found that fish were reliably recruited by the mobile dummy if it emitted electrical signals and followed it largely independently of the presented playback patterns. While following the dummy, fish interacted with it spatially by displaying stereotypical motor patterns, as well as electrically, e.g. through discharge regularizations and by synchronizing their own discharge activity to the playback. However, the overall emission frequencies of the dummy were not adopted by the following fish. Instead, social signals based on different temporal patterns were emitted depending on the type of playback. In particular, double pulses were displayed in response to electrical signaling of the dummy and their expression was positively correlated with an animals' rank in the dominance hierarchy. Based on additional analysis of swimming trajectories and stereotypical locomotor behavior patterns, we conclude that the reception and emission of electrical communication signals play a crucial role in mediating social interactions in mormyrid weakly electric fish. PMID:28902915

No effect of bipolar interferential electrotherapy and pulsed ultrasound for soft tissue shoulder disorders: a randomised controlled trial

PubMed Central

van der Heijden, G. J M G; Leffers, P.; Wolters, P.; Verheijden, J.; van Mameren, H.; Houben, J.; Bouter, L.; Knipschild, P.

1999-01-01

OBJECTIVE—To assess the efficacy of bipolar interferential electrotherapy (ET) and pulsed ultrasound (US) as adjuvants to exercise therapy for soft tissue shoulder disorders (SD).
METHODS—Randomised placebo controlled trial with a two by two factorial design plus an additional control group in 17 primary care physiotherapy practices in the south of the Netherlands. Patients with shoulder pain and/or restricted shoulder mobility, because of a soft tissue impairment without underlying specific or generalised condition, were enrolled if they had not recovered after six sessions of exercise therapy in two weeks. They were randomised to receive (1) active ET plus active US; (2) active ET plus dummy US; (3) dummy ET plus active US; (4) dummy ET plus dummy US; or (5) no adjuvants. Additionally, they received a maximum of 12 sessions of exercise therapy in six weeks. Measurements at baseline, 6 weeks and 3, 6, 9, and 12 months later were blinded for treatment. Outcome measures: recovery, functional status, chief complaint, pain, clinical status, and range of motion.
RESULTS—After written informed consent 180 patients were randomised: both the active treatments were given to 73 patients, both the dummy treatments to 72 patients, and 35 patients received no adjuvants. Prognosis of groups appeared similar at baseline. Blinding was successfully maintained. At six weeks seven patients (20%) without adjuvants reported very large improvement (including complete recovery), 17 (23%) and 16 (22%) with active and dummy ET, and 19 (26%) and 14 (19%) with active and dummy US. These proportions increased to about 40% at three months, but remained virtually stable thereafter. Up to 12 months follow up the 95% CI for differences between groups for all outcomes include zero.
CONCLUSION—Neither ET nor US prove to be effective as adjuvants to exercise therapy for soft tissue SD.

 PMID:10460185

Beam heated linear theta-pinch device for producing hot plasmas

DOEpatents

Bohachevsky, Ihor O.

1981-01-01

A device for producing hot plasmas comprising a single turn theta-pinch coil, a fast discharge capacitor bank connected to the coil, a fuel element disposed along the center axis of the coil, a predetermined gas disposed within the theta-pinch coil, and a high power photon, electron or ion beam generator concentrically aligned to the theta-pinch coil. Discharge of the capacitor bank generates a cylindrical plasma sheath within the theta-pinch coil which heats the outer layer of the fuel element to form a fuel element plasma layer. The beam deposits energy in either the cylindrical plasma sheath or the fuel element plasma layer to assist the implosion of the fuel element to produce a hot plasma.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

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

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

49 CFR 572.131 - General description.

Code of Federal Regulations, 2014 CFR

2014-10-01

... Female Test Dummy, Alpha Version § 572.131 General description. (a) The Hybrid III fifth percentile adult... Small Adult Female Crash Test Dummy (HIII-5F, Alpha Version) (June 2002) (refer to § 572.130(a)(1)(ix...

49 CFR 572.131 - General description.

Code of Federal Regulations, 2010 CFR

2010-10-01

... Female Test Dummy, Alpha Version § 572.131 General description. (a) The Hybrid III fifth percentile adult... Small Adult Female Crash Test Dummy (HIII-5F, Alpha Version) (June 2002) (refer to § 572.130(a)(1)(ix...

49 CFR 572.131 - General description.

Code of Federal Regulations, 2012 CFR

2012-10-01

... Female Test Dummy, Alpha Version § 572.131 General description. (a) The Hybrid III fifth percentile adult... Small Adult Female Crash Test Dummy (HIII-5F, Alpha Version) (June 2002) (refer to § 572.130(a)(1)(ix...

49 CFR 572.121 - General description.

Code of Federal Regulations, 2010 CFR

2010-10-01

... Dummy, Beta Version § 572.121 General description. (a) The Hybrid III type 6-year-old dummy is defined... specifications package P/N 127-0000, the titles of which are listed in Table A; (2) Procedures for Assembly...

BOILER-SUPERHEATED REACTOR

DOEpatents

Heckman, T.P.

1961-05-01

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-11-21

A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

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

Pope, M. A.; DeHart, M. D.; Morrell, S. R.

2015-03-01

Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

DOEpatents

Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

1982-01-01

A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

Quality assurance in MR image guided adaptive brachytherapy for cervical cancer: Final results of the EMBRACE study dummy run.

PubMed

Kirisits, Christian; Federico, Mario; Nkiwane, Karen; Fidarova, Elena; Jürgenliemk-Schulz, Ina; de Leeuw, Astrid; Lindegaard, Jacob; Pötter, Richard; Tanderup, Kari

2015-12-01

Upfront quality assurance (QA) is considered essential when starting a multicenter clinical trial in radiotherapy. Despite the long experience gained for external beam radiotherapy (EBRT) trials, there are only limited audit QA methods for brachytherapy (BT) and none include the specific aspects of image guided adaptive brachytherapy (IGABT). EMBRACE is a prospective multicenter trial aiming to assess the impact of (MRI)-based IGABT in locally advanced cervical cancer. An EMBRACE dummy run was designed to identify sources and magnitude of uncertainties and errors considered important for the evaluation of clinical, and dosimetric parameters and their relation to outcome. Contouring, treatment planning and dose reporting was evaluated and scored with a categorical scale of 1-10. Active feedback to centers was provided to improve protocol compliance and reporting. A second dummy run was required in case of major deviations (score <7) for any item. Overall 27/30 centers passed the dummy run. 16 centers had to repeat the dummy run in order to clarify major inconsistencies to the protocol. The most pronounced variations were related to contouring for both EBRT and BT. Centers with experience in IGABT (>30 cases) had better performance as compared to centers with limited experience. The comprehensive dummy run designed for the EMBRACE trial has been a feasible tool for QA in IGABT of cervix cancer. It should be considered for future IGABT trials and could serve as the basis for continuous quality checks for brachytherapy centers. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

Head Excursion of Restrained Human Volunteers and Hybrid III Dummies in Steady State Rollover Tests

PubMed Central

Moffatt, Edward; Hare, Barry; Hughes, Raymond; Lewis, Lance; Iiyama, Hiroshi; Curzon, Anne; Cooper, Eddie

2003-01-01

Seatbelts provide substantial benefits in rollover crashes, yet occupants still receive head and neck injuries from contacting the vehicle roof interior when the roof exterior strikes the ground. Prior research has evaluated rollover restraint performance utilizing anthropomorphic test devices (dummies), but little dynamic testing has been done with human volunteers to learn how they move during rollovers. In this study, the vertical excursion of the head of restrained dummies and human subjects was measured in a vehicle being rotated about its longitudinal roll axis at roll rates from 180-to-360 deg/sec and under static inversion conditions. The vehicle’s restraint design was the commonly used 3-point seatbelt with continuous loop webbing and a sliding latch plate. This paper presents an analysis of the observed occupant motion and provides a comparison of dummy and human motion under similar test conditions. Thirty-five tests (eighteen static and seventeen dynamic) were completed using two different sizes of dummies and human subjects in both near and far-side roll directions. The research indicates that far-side rollovers cause the restrained test subjects to have greater head excursion than near-side rollovers, and that static inversion testing underestimates head excursion for far-side occupants. Human vertical head excursion of up to 200 mm was found at a roll rate of 220 deg/sec. Humans exhibit greater variability in head excursion in comparison to dummies. Transfer of seatbelt webbing through the latch plate did not correlate directly with differences in head excursion. PMID:12941241

Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Guba, G. G.

2017-11-01

This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

Techniques for Developing Child Dummy Protection Reference Values. Event Report

DOT National Transportation Integrated Search

1996-10-01

The purpose of this report is to present background information and techniques : for developing protection reference values (PRV) to use with child dummies in : out-of-position (OOP) child/air bag interaction testing. Biomechanics experts : agree tha...

THE MANUFACTURE OF FUEL ELEMENTS OF THE ARGONAUT TYPE

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

Kittl, J.; Machado, R.E.; Mazza, J.A.

1958-06-10

The conditions required for the manufacture of the RA-1 Argonant type fuel elements are investigated. The fuel elements are in the form of a plate which is manufactured by the extrusion of a presintered mass of U/sub 3/O/sub 8/ (20% enriched) in an aluminum matrix. Steps in the investigation were obtention and specification of U/sub 3/O/sub 8/ and Al in powder form for testing, filling, and extrusion tests, finishing of the fuel elements, and computation of U/sub 3/O/sub 8/ content. (W.D.M.)

Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

NASA Technical Reports Server (NTRS)

Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

2013-01-01

In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

40 CFR 79.56 - Fuel and fuel additive grouping system.

Code of Federal Regulations, 2010 CFR

2010-07-01

... further testing under the provisions of Tier 3 or to support regulatory decisions affecting that fuel or... elements or classes of compounds other than those permitted in the base fuel for the respective fuel family... all of the following criteria: (1) Contain no elements other than carbon, hydrogen, oxygen, nitrogen...

Techniques for developing child dummy protection reference values : event report

DOT National Transportation Integrated Search

1996-10-01

The purpose of this report is to present background information and techniques for developing protection reference values (PRV) to use with child dummies in out-of-position (OOP) child/air bag interaction testing. This report summarizes the literatur...

Potential technique for improving the survival of victims of tsunamis

PubMed Central

Suga, Hisami; Prochazka, Zdenek; Suzuki, Kojiro; Oguri, Kazumasa; Inoue, Tetsunori

2018-01-01

We investigated a method for surviving tsunamis that involved the use of personal flotation devices (PFDs). In our work, we succeeded in numerically demonstrating that the heads of all the dummies wearing PFDs remained on the surface and were not dragged underwater after the artificial tsunami wave hit them. In contrast, the heads of all the dummies not wearing PFDs were drawn underwater immediately; these dummies were subsequently entrapped in a vortex. The results of our series of experiments are important as a first step to preventing the tragedies caused by tsunamis. PMID:29791490

MULTIPLE SETS OF TWIN SLABS ON THE RUN OUT. THE ...

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

MULTIPLE SETS OF TWIN SLABS ON THE RUN OUT. THE RUN OUT INCLUDES THE TRAVELING TORCH WHICH CUTS SLABS TO DESIRED LENGTH, AN IDENTIFICATION SYSTEM TO INDICATE HEAT NUMBER AND TRACE IDENTITY OF EVERY SLAB, AND A DEBURRING DEVICE TO SMOOTH SLABS. AT LEFT OF ROLLS IS THE DUMMY BAR. DUMMY BAR IS INSERTED UP THROUGH CONTAINMENT SECTION INTO MOLD PRIOR TO START OF CAST. WHEN STEEL IS INTRODUCED INTO MOLD IT CONNECTS WITH BAR AS CAST BEGINS, AT RUN OUT DUMMY BAR DISCONNECTS AND IS STORED. - U.S. Steel, Fairfield Works, Continuous Caster, Fairfield, Jefferson County, AL

MULTIPLE SETS OF TWIN SLABS ON THE RUN OUT. THE ...

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

MULTIPLE SETS OF TWIN SLABS ON THE RUN OUT. THE RUN OUT INCLUDES THE TRAVELING TORCH WHICH CUTS SLABS TO DESIRED LENGTH, AN IDENTIFICATION SYSTEM TO INDICATE HEAT NUMBER AND TRACE IDENTITY OF EVERY SLAB, AND A DEBURRING DEVICE TO SMOOTH SLABS. AT LEFT OF ROLLS IS THE DUMMY BAR. DUMMY BAR IS INSERTED UP THROUGH CONTAINMENT SECTION INTO MOLD PRIOR TO START OF CAST. WHEN STEEL IS INTRODUCED INTO MOLD IT CONNECTS WITH BAR AS CAST BEGINS, AT RUN OUT DUMMY BAR DISCONNECTS AND IS STORED - U.S. Steel, Fairfield Works, Continuous Caster, Fairfield, Jefferson County, AL

Comparison of car seats in low speed rear-end impacts using the BioRID dummy and the new neck injury criterion (NIC).

PubMed

Boström, O; Fredriksson, R; Håland, Y; Jakobsson, L; Krafft, M; Lövsund, P; Muser, M H; Svensson, M Y

2000-03-01

Long-term whiplash associated disorders (WAD) 1-3 sustained in low velocity rear-end impacts is the most common disability injury in Sweden. Therefore, to determine neck injury mechanisms and develop methods to measure neck-injury related parameters are of importance for current crash-safety research. A new neck injury criterion (NIC) has previously been proposed and evaluated by means of dummy, human and mathematical rear-impact simulations. So far, the criterion appears to be sensitive to the major car and collision related risk factors for injuries with long-term consequences. To further evaluate the applicability of NIC, four seats were tested according to a recently proposed sled-test procedure. 'Good' as well as 'bad' seats were chosen on the basis of a recently presented disability risk ranking list. The dummy used in the current tests was the Biofidelic Rear Impact Dummy (BioRID). The results of this study showed that NICmax values were generally related to the real-world risk of long-term WAD 1-3. Furthermore, these results suggested that NICmax calculated from sled tests using the BioRID dummy can be used for evaluating the neck injury risk of different car seats.

Benefits of a Low Severity Frontal Crash Test

PubMed Central

Digges, Kennerly; Dalmotas, Dainius

2007-01-01

The US Federal Motor Vehicle Safety Standard for frontal protection requires vehicle crash tests into a rigid barrier with two belted dummies in the front seats. The standard was recently modified to require two separate 56 Kph frontal tests. In one test the dummies are 50% males. In the other test, the dummies are 5% females. Analysis of crash test data indicates that the 56 Kph test does not encourage technology to reduce chest injuries in lower severity crashes. Tests conducted by Transport Canada provide data from belted 5% female dummies in the front seats of vehicles that were subjected crashes into a rigid barrier at 40 Kph. An analysis of the results showed that for many vehicles, the risks of serious chest injuries were higher in the 40 Kph test than in a 56 Kph test. This paper examines the benefits that would result from a requirement for a low severity (40 Kph) frontal barrier crash test with two belted 5% female dummies and more stringent chest injury requirements. A preliminary benefits analysis for chest deflection allowable in the range of 28 mm. to 36 mm. was conducted. A standard that limits the chest deflection to 34 mm. would reduce serious chest injury by 16% to 24% for the belted population in frontal crashes. PMID:18184499

Benefits of a low severity frontal crash test.

PubMed

Digges, Kennerly; Dalmotas, Dainius

2007-01-01

The US Federal Motor Vehicle Safety Standard for frontal protection requires vehicle crash tests into a rigid barrier with two belted dummies in the front seats. The standard was recently modified to require two separate 56 Kph frontal tests. In one test the dummies are 50% males. In the other test, the dummies are 5% females. Analysis of crash test data indicates that the 56 Kph test does not encourage technology to reduce chest injuries in lower severity crashes. Tests conducted by Transport Canada provide data from belted 5% female dummies in the front seats of vehicles that were subjected crashes into a rigid barrier at 40 Kph. An analysis of the results showed that for many vehicles, the risks of serious chest injuries were higher in the 40 Kph test than in a 56 Kph test. This paper examines the benefits that would result from a requirement for a low severity (40 Kph) frontal barrier crash test with two belted 5% female dummies and more stringent chest injury requirements. A preliminary benefits analysis for chest deflection allowable in the range of 28 mm. to 36 mm. was conducted. A standard that limits the chest deflection to 34 mm. would reduce serious chest injury by 16% to 24% for the belted population in frontal crashes.

Refined Dummy Atom Model of Mg(2+) by Simple Parameter Screening Strategy with Revised Experimental Solvation Free Energy.

PubMed

Jiang, Yang; Zhang, Haiyang; Feng, Wei; Tan, Tianwei

2015-12-28

Metal ions play an important role in the catalysis of metalloenzymes. To investigate metalloenzymes via molecular modeling, a set of accurate force field parameters for metal ions is highly imperative. To extend its application range and improve the performance, the dummy atom model of metal ions was refined through a simple parameter screening strategy using the Mg(2+) ion as an example. Using the AMBER ff03 force field with the TIP3P model, the refined model accurately reproduced the experimental geometric and thermodynamic properties of Mg(2+). Compared with point charge models and previous dummy atom models, the refined dummy atom model yields an enhanced performance for producing reliable ATP/GTP-Mg(2+)-protein conformations in three metalloenzyme systems with single or double metal centers. Similar to other unbounded models, the refined model failed to reproduce the Mg-Mg distance and favored a monodentate binding of carboxylate groups, and these drawbacks needed to be considered with care. The outperformance of the refined model is mainly attributed to the use of a revised (more accurate) experimental solvation free energy and a suitable free energy correction protocol. This work provides a parameter screening strategy that can be readily applied to refine the dummy atom models for metal ions.

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

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

Sell, D. A.; Baily, C. E.; Malewitz, T. J.

2016-09-01

A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium aftermore » the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.« less

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.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

Nuclear fuel elements having a composite cladding

DOEpatents

Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

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

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

METHOD AND APPARATUS FOR HANDLING RADIOACTIVE PRODUCTS

DOEpatents

Nicoll, D.

1959-02-24

A device is described for handling fuel elements being discharged from a nuclear reactor. The device is adapted to be disposed beneath a reactor within the storage canal for spent fuel elements. The device is comprised essentially of a cylinder pivotally mounted to a base for rotational motion between a vertical position. where the mouth of the cylinder is in the top portion of the container for receiving a fuel element discharged from a reactor into the cylinder, and a horizontal position where the mouth of the cylinder is remote from the top portion of the container and the fuel element is discharged from the cylinder into the storage canal. The device is operated by hydraulic pressure means and is provided with a means to prevent contaminated primary liquid coolant in the reactor system from entering the storage canal with the spent fuel element.

Characteristics of the Injury Environment in Far-Side Crashes

PubMed Central

Digges, K.; Gabler, H; Mohan, P.; Alonso, B.

2005-01-01

The population of occupants in far-side crashes that are documented in the US National database (NASS/CDS) was studied. The annual number of front seat occupants with serious or fatal injuries in far-side planar and rollover crashes was 17,194. The crash environment that produces serious and fatal injuries to belted front seat occupants in planar far-side crashes was investigated in detail. It was found that both the change in velocity and extent of damage were important factors that relate to crash severity. The median severity for crashes with serious or fatal injuries was a lateral delta-V of 28 kph and an extent of damage of CDC 3.6. Vehicle-to-vehicle impacts were simulated by finite element models to determine the intrusion characteristics associated with the median crash condition. These simulations indicated that the side damage caused by the IIHS barrier was representative of the damage in crashes that produce serious injuries in far-side crashes. Occupant simulations of the IIHS barrier crash at 28 kph showed that existing dummies lack biofidelity in upper body motion. The analysis suggested test conditions for studying far-side countermeasures and supported earlier studies that showed the need for an improved dummy to evaluate safety performance in the far-side crash environment. PMID:16179148

Calculation of Distribution Dynamics of Inhomogeneous Temperature Field in Range of Fuel Elements by Using FreeFem++

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Shishkin, A. V.

2017-11-01

This article introduces the result of studying the heat exchange in the fuel element of the nuclear reactor fuel magazine. Fuel assemblies are completed as a bundle of cylindrical fuel elements located at the tops of a regular triangle. Uneven distribution of fuel rods in a nuclear reactor’s core forms the inhomogeneity of temperature fields. This article describes the developed method for heat exchange calculation with the account for impact of an inhomogeneous temperature field on the thermal-physical properties of materials and unsteady effects. The acquired calculation results are used for evaluating the tolerable temperature levels in protective case materials.

Fuel cell elements with improved water handling capacity

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor); Lee, Albany (Inventor)

2001-01-01

New fuel cell components for use in liquid feed fuel cell systems are provided. The components include biplates and endplates, having a hydrophilic surface and allow high efficiency operation. Conductive elements and a wicking device also form a part of the fuel cell components of the invention.

Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

NASA Astrophysics Data System (ADS)

Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

2017-04-01

Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

49 CFR 572.31 - General description.

Code of Federal Regulations, 2010 CFR

2010-10-01

..., titled “Sign Convention for Vehicle Crash Testing”, dated 1994-12. (6) Exterior dimensions of the Hybrid... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES Hybrid III Test Dummy § 572.31 General description. (a) The Hybrid III 50th percentile size dummy consists of components and...

49 CFR 572.141 - General description.

Code of Federal Regulations, 2010 CFR

2010-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES 3-year-Old Child Crash Test Dummy, Alpha Version § 572.141 General description. (a) The Hybrid III 3-year-old child dummy is described by the following materials: (1) Technical drawings and specifications package 210-0000 (refer to...

EFFECTS OF OVERPRESSURES IN GROUP SHELTERS ON ANIMALS AND DUMMIES

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

Roberts, J.E.; White, C.S.; Chiffelle, T.L.

1953-09-01

S>Relative biological hazards of blast were studied in two types of communal air-raid shelters during Shots 1 and 8. Dogs, restrained within the shelters during detonation, were studied pathologically and clinically for blast injuries. Two anthropometric dummies were test objects for displacement studies utilizing high-speed photography. Physical data included pressure vs time and air-drag determinations. During Shot 1, animals sustained marked blast damages (hemorrhages in lungs and abdominal organs), three dogs were ataxic. and the dummies were rather violently displaced. In Shot 8, however, no significant injuries were found in the animals, and the dummies were minimally displaced. Analysis ofmore » the physical data indicated that blast injuries and violent displacements may occur at much lower static overpressures than previously assumed from conventional explosion data. Furthermore, biological damage appeared to be related to the rate of rise of the overpressure and air drag, as well as the maximum overpressure values. (auth)« less

Magnetic dummy molecularly imprinted polymers based on multi-walled carbon nanotubes for rapid selective solid-phase extraction of 4-nonylphenol in aqueous samples.

PubMed

Rao, Wei; Cai, Rong; Yin, Yuli; Long, Fang; Zhang, Zhaohui

2014-10-01

In this paper, a highly selective sample clean-up procedure combining magnetic dummy molecular imprinting with solid-phase extraction was developed for rapid separation and determination of 4-nonylphenol (NP) in the environmental water samples. The magnetic dummy molecularly imprinted polymers (mag-DMIPs) based on multi-walled carbon nanotubes were successfully synthesized with a surface molecular imprinting technique using 4-tert-octylphenol as the dummy template and tetraethylorthosilicate as the cross-linker. The maximum adsorption capacity of the mag-DMIPs for NP was 52.4 mg g(-1) and it took about 20 min to achieve the adsorption equilibrium. The mag-DMIPs exhibited the specific selective adsorption toward NP. Coupled with high performance liquid chromatography analysis, the mag-DMIPs were used to extract solid-phase and detect NP in real water samples successfully with the recoveries of 88.6-98.1%. Copyright © 2014 Elsevier B.V. All rights reserved.

78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.

1957-10-01

A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

Transformer current sensor for superconducting magnetic coils

DOEpatents

Shen, S.S.; Wilson, C.T.

1985-04-16

The present invention is a current transformer for operating currents larger than 2kA (two kiloamps) that is capable of detecting a millivolt level resistive voltage in the presence of a large inductive voltage. Specifically, the present invention includes substantially cylindrical primary turns arranged to carry a primary current and substantially cylindrical secondary turns arranged coaxially with and only partially within the primary turns, the secondary turns including an active winding and a dummy winding, the active and dummy windings being coaxial, longitudinally separated and arranged to mutually cancel voltages excited by commonly experienced magnetic fields, the active winding but not the dummy winding being arranged within the primary turns.

Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2017-01-01

To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

Summary of new test dummies, injury criteria

DOT National Transportation Integrated Search

2000-06-17

Besides a plethora of new tests, the air bag standard issued recently calls for four new test dummies in addition to the average size adult male already used in testing: small adult female; 6 year old child; 3 year old child; and 1 year old infant. I...

Using Time-Series Regression to Predict Academic Library Circulations.

ERIC Educational Resources Information Center

Brooks, Terrence A.

1984-01-01

Four methods were used to forecast monthly circulation totals in 15 midwestern academic libraries: dummy time-series regression, lagged time-series regression, simple average (straight-line forecasting), monthly average (naive forecasting). In tests of forecasting accuracy, dummy regression method and monthly mean method exhibited smallest average…

Space shuttle orbit maneuvering engine, reusable thrust chamber program. Task 6: Data dump hot fuel element investigation

NASA Technical Reports Server (NTRS)

Nurick, W. H.

1974-01-01

An evaluation of reusable thrust chambers for the space shuttle orbit maneuvering engine was conducted. Tests were conducted using subscale injector hot-fire procedures for the injector configurations designed for a regenerative cooled engine. The effect of operating conditions and fuel temperature on combustion chamber performance was determined. Specific objectives of the evaluation were to examine the optimum like-doublet element geometry for operation at conditions consistent with a fuel regeneratively cooled engine (hot fuel, 200 to 250 F) and the sensitivity of the triplet injector element to hot fuels.

Tag gas capsule with magnetic piercing device

DOEpatents

Nelson, Ira V.

1976-06-22

An apparatus for introducing a tag (i.e., identifying) gas into a tubular nuclear fuel element. A sealed capsule containing the tag gas is placed in the plenum in the fuel tube between the fuel and the end cap. A ferromagnetic punch having a penetrating point is slidably mounted in the plenum. By external electro-magnets, the punch may be caused to penetrate a thin rupturable end wall of the capsule and release the tag gas into the fuel element. Preferably the punch is slidably mounted within the capsule, which is in turn loaded as a sealed unit into the fuel element.

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Nuclear reactor fuel element having improved heat transfer

DOEpatents

Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

1982-03-03

A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

DOEpatents

Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

1957-11-12

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

MRT fuel element inspection at Dounreay

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

Gibson, J.

1997-08-01

To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

Direct carbon fuel cell and stack designs

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

Gorte, Raymond J.; Oh, Tae-Sik

Disclosed are novel configurations of Direct Carbon Fuel Cells (DCFCs), which optionally comprise a liquid anode. The liquid anode comprises a molten salt/metal, preferably Sb, and a fuel, which has significant elemental carbon content (coal, bio-mass, etc.). The supply of fuel is continuously replenished in the anode. In addition, a stack configuration is suggested where combining a large number of planar or tubular fuel elements.

Nuclear breeder reactor fuel element with axial tandem stacking and getter

DOEpatents

Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

1981-01-01

A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

Leadership for Dummies: A Capstone Project for Leadership Students

ERIC Educational Resources Information Center

Moore, Lori L.; Odom, Summer F.; Wied, Lexi M.

2011-01-01

Capstone courses in leadership provide students opportunities to synthesize prior knowledge about various aspects of leadership. This article describes the "Leadership for Dummies" project, which could be used as a capstone experience for leadership majors. Based on his experiences as a psychological researcher, Gardner (2008) identified five…

Regression Analysis with Dummy Variables: Use and Interpretation.

ERIC Educational Resources Information Center

Hinkle, Dennis E.; Oliver, J. Dale

1986-01-01

Multiple regression analysis (MRA) may be used when both continuous and categorical variables are included as independent research variables. The use of MRA with categorical variables involves dummy coding, that is, assigning zeros and ones to levels of categorical variables. Caution is urged in results interpretation. (Author/CH)

49 CFR 572.181 - General description.

Code of Federal Regulations, 2012 CFR

2012-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES 2re Side Impact Crash Test Dummy, 50th Percentile Adult Male § 572.181 General description. (a) The ES-2re Side Impact Crash Test... (PADI) of the ES-2re Side Impact Crash Test Dummy, February 2008, incorporated by reference, see § 572...

49 CFR 572.181 - General description.

Code of Federal Regulations, 2013 CFR

2013-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES ES-2re Side Impact Crash Test Dummy, 50th Percentile Adult Male § 572.181 General description. (a) The ES-2re Side Impact Crash... (PADI) of the ES-2re Side Impact Crash Test Dummy, February 2008, incorporated by reference, see § 572...

49 CFR 572.191 - General description.

Code of Federal Regulations, 2013 CFR

2013-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES SID-IIsD Side Impact Crash Test Dummy, Small Adult Female § 572.191 General description. (a) The SID-IIsD Side Impact Crash Test... test sensors for the SID-IIsD Side Impact Crash Test Dummy, 5th percentile adult female, is shown in...

49 CFR 572.181 - General description.

Code of Federal Regulations, 2014 CFR

2014-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES ES-2re Side Impact Crash Test Dummy, 50th Percentile Adult Male § 572.181 General description. (a) The ES-2re Side Impact Crash... (PADI) of the ES-2re Side Impact Crash Test Dummy, February 2008, incorporated by reference, see § 572...

49 CFR 572.191 - General description.

Code of Federal Regulations, 2014 CFR

2014-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) ANTHROPOMORPHIC TEST DEVICES SID-IIsD Side Impact Crash Test Dummy, Small Adult Female § 572.191 General description. (a) The SID-IIsD Side Impact Crash Test... test sensors for the SID-IIsD Side Impact Crash Test Dummy, 5th percentile adult female, is shown in...

49 CFR 572.120 - Incorporation by reference.

Code of Federal Regulations, 2010 CFR

2010-10-01

... Child Test Dummy, Beta Version § 572.120 Incorporation by reference. (a) The following materials are... List and Drawings, Hybrid III Six-year-old Child Test Dummy (H-III6C, Beta Version) (June 2002... (vii) The Hybrid III Six-year-old Child Parts/Drawing List. (2) A procedures manual entitled...

Segmented Polynomial Models in Quasi-Experimental Research.

ERIC Educational Resources Information Center

Wasik, John L.

1981-01-01

The use of segmented polynomial models is explained. Examples of design matrices of dummy variables are given for the least squares analyses of time series and discontinuity quasi-experimental research designs. Linear combinations of dummy variable vectors appear to provide tests of effects in the two quasi-experimental designs. (Author/BW)

Two-Dimensional Diffusion Theory Analysis of Reactivity Effects of a Fuel-Plate-Removal Experiment

NASA Technical Reports Server (NTRS)

Gotsky, Edward R.; Cusick, James P.; Bogart, Donald

1959-01-01

Two-dimensional two-group diffusion calculations were performed on the NASA reactor simulator in order to evaluate the reactivity effects of fuel plates removed successively from the center experimental fuel element of a seven- by three-element core loading at the Oak Ridge Bulk Shielding Facility. The reactivity calculations were performed by two methods: In the first, the slowing-down properties of the experimental fuel element were represented by its infinite media parameters; and, in the second, the finite size of the experimental fuel element was recognized, and the slowing-down properties of the surrounding core were attributed to this small region. The latter calculation method agreed very well with the experimented reactivity effects; the former method underestimated the experimental reactivity effects.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

Toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Priest, N D; Richardson, R B; Edwards, G W R

2013-02-01

The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

Nuclear fuel pin scanner

DOEpatents

Bramblett, Richard L.; Preskitt, Charles A.

1987-03-03

Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

REACTOR FUEL ELEMENTS TESTING CONTAINER

DOEpatents

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Determination of trace elements in automotive fuels by filter furnace atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Anselmi, Anna; Tittarelli, Paolo; Katskov, Dmitri A.

2002-03-01

The determination of Cd, Cr, Cu, Pb and Ni was performed in gasoline and diesel fuel samples by electrothermal atomic absorption spectrometry using the Transverse Heated Filter Atomizer (THFA). Thermal conditions were experimentally defined for the investigated elements. The elements were analyzed without addition of chemical modifiers, using organometallic standards for the calibration. Forty-microliter samples were injected into the THFA. Gasoline samples were analyzed directly, while diesel fuel samples were diluted 1:4 with n-heptane. The following characteristic masses were obtained: 0.8 pg Cd, 6.4 pg Cr, 12 pg Cu, 17 pg Pb and 27 pg Ni. The limits of determination for gasoline samples were 0.13 μg/kg Cd, 0.4 μg/kg Cr, 0.9 μg/kg Cu, 1.5 μg/kg Pb and 2.5 μg/kg Ni. The corresponding limit of determination for diesel fuel samples was approximately four times higher for all elements. The element recovery was performed using the addition of organometallic compounds to gasoline and diesel fuel samples and was between 85 and 105% for all elements investigated.

FUEL ELEMENT

DOEpatents

Bean, R.W.

1963-11-19

A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

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

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

Rest, J.; Hofman, G.L.

1997-12-01

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

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.

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

16 CFR 1216.2 - Requirements for infant walkers.

Code of Federal Regulations, 2011 CFR

2011-01-01

... coefficient of friction = 0.05 NCAMI = Normal force (for CAMI dummy scenario) = weight of CAMI dummy and... occupant seating area and arms placed on the walker tray. (ii) [Reserved] (8) Instead of complying with... horizontally (0 ± 0.5° with respect to the table surface). (ii) [Reserved] (9) Instead of complying with...

49 CFR 572.197 - Abdomen.

Code of Federal Regulations, 2010 CFR

2010-10-01

... vertical orientation. (4) Push the dummy at the knees and at mid-sternum of the upper torso with just sufficient horizontally oriented force towards the seat back until the back of the upper torso is in contact with the seat back. (5) While maintaining the dummy's position as specified in paragraph (b)(3) and (4...

49 CFR 572.196 - Thorax without arm.

Code of Federal Regulations, 2010 CFR

2010-10-01

... orientation. (4) Push the dummy at the knees and at mid-sternum of the upper torso with just sufficient horizontally oriented force towards the seat back until the back of the upper torso is in contact with the seat back. (5) While maintaining the dummy's position as specified in paragraphs (b)(3) and (4) of this...

49 CFR 572.186 - Abdomen assembly.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 7 2013-10-01 2013-10-01 false Abdomen assembly. 572.186 Section 572.186... Test Dummy, 50th Percentile Adult Male § 572.186 Abdomen assembly. (a) The abdomen assembly (175-5000) is part of the dummy assembly shown in drawing 175-0000 including load sensors specified in § 572.189...

49 CFR 572.186 - Abdomen assembly.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 7 2011-10-01 2011-10-01 false Abdomen assembly. 572.186 Section 572.186... Dummy, 50th Percentile Adult Male § 572.186 Abdomen assembly. (a) The abdomen assembly (175-5000) is part of the dummy assembly shown in drawing 175-0000 including load sensors specified in § 572.189(e...

49 CFR 572.186 - Abdomen assembly.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 7 2012-10-01 2012-10-01 false Abdomen assembly. 572.186 Section 572.186... Dummy, 50th Percentile Adult Male § 572.186 Abdomen assembly. (a) The abdomen assembly (175-5000) is part of the dummy assembly shown in drawing 175-0000 including load sensors specified in § 572.189(e...

49 CFR 571.213 - Standard No. 213; Child restraint systems.

Code of Federal Regulations, 2012 CFR

2012-10-01

... the initial pre-test position of the respective knee pivot point, measured along a horizontal line... test dummy, specified in S7, when a child restraint system is tested in accordance with S6.1. Factory... body of a seated anthropomorphic test dummy, excluding the thighs, that lies between the top of the...

49 CFR 571.213 - Standard No. 213; Child restraint systems.

Code of Federal Regulations, 2013 CFR

2013-10-01

... the initial pre-test position of the respective knee pivot point, measured along a horizontal line... test dummy, specified in S7, when a child restraint system is tested in accordance with S6.1. Factory... body of a seated anthropomorphic test dummy, excluding the thighs, that lies between the top of the...

49 CFR 571.213 - Standard No. 213; Child restraint systems.

Code of Federal Regulations, 2014 CFR

2014-10-01

... the initial pre-test position of the respective knee pivot point, measured along a horizontal line... the head or torso of the appropriate test dummy, specified in S7, when a child restraint system is... (§ 571.225). Torso means the portion of the body of a seated anthropomorphic test dummy, excluding the...

49 CFR 572.196 - Thorax without arm.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 7 2012-10-01 2012-10-01 false Thorax without arm. 572.196 Section 572.196... Dummy, Small Adult Female § 572.196 Thorax without arm. (a) The thorax is part of the upper torso... (drawing 180-0000) with the arm (180-6000) on the impacted side removed. The dummy's thorax is equipped...

49 CFR 572.196 - Thorax without arm.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 7 2014-10-01 2014-10-01 false Thorax without arm. 572.196 Section 572.196... Test Dummy, Small Adult Female § 572.196 Thorax without arm. (a) The thorax is part of the upper torso... (drawing 180-0000) with the arm (180-6000) on the impacted side removed. The dummy's thorax is equipped...

49 CFR 572.196 - Thorax without arm.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 7 2011-10-01 2011-10-01 false Thorax without arm. 572.196 Section 572.196... Dummy, Small Adult Female § 572.196 Thorax without arm. (a) The thorax is part of the upper torso... (drawing 180-0000) with the arm (180-6000) on the impacted side removed. The dummy's thorax is equipped...

49 CFR 572.196 - Thorax without arm.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 7 2013-10-01 2013-10-01 false Thorax without arm. 572.196 Section 572.196... Test Dummy, Small Adult Female § 572.196 Thorax without arm. (a) The thorax is part of the upper torso... (drawing 180-0000) with the arm (180-6000) on the impacted side removed. The dummy's thorax is equipped...

49 CFR 572.150 - Incorporation by reference.

Code of Federal Regulations, 2014 CFR

2014-10-01

...-Month-Old Infant, Alpha Version § 572.150 Incorporation by reference. (a) The following materials are... Drawings, Subpart R, CRABI 12-Month-Old Infant Crash Test Dummy (CRABI-12, Alpha version) August 2001” and... Infant Crash Test Dummy (CRABI-12, Alpha version) August 2001” incorporated by reference in § 572.155; (3...

49 CFR 572.150 - Incorporation by reference.

Code of Federal Regulations, 2013 CFR

2013-10-01

...-Month-Old Infant, Alpha Version § 572.150 Incorporation by reference. (a) The following materials are... Drawings, Subpart R, CRABI 12-Month-Old Infant Crash Test Dummy (CRABI-12, Alpha version) August 2001” and... Infant Crash Test Dummy (CRABI-12, Alpha version) August 2001” incorporated by reference in § 572.155; (3...

49 CFR 572.150 - Incorporation by reference.

Code of Federal Regulations, 2012 CFR

2012-10-01

...-Month-Old Infant, Alpha Version § 572.150 Incorporation by reference. (a) The following materials are... Drawings, Subpart R, CRABI 12-Month-Old Infant Crash Test Dummy (CRABI-12, Alpha version) August 2001” and... Infant Crash Test Dummy (CRABI-12, Alpha version) August 2001” incorporated by reference in § 572.155; (3...

49 CFR 572.151 - General description.

Code of Federal Regulations, 2010 CFR

2010-10-01

... specifications package 921022-000 (refer to § 572.150(a)(1)), the titles of which are listed in Table A of this...)). (b) The dummy consists of the component assemblies set out in the following Table A: Table A... dummy are joined in a manner such that, except for contacts existing under static conditions, there is...

49 CFR 572.43 - Lumbar spine and pelvis.

Code of Federal Regulations, 2010 CFR

2010-10-01

... vertical plane which is tangent to the back of the dummy's buttocks. (3) Align the test probe so that at... vertical planes perpendicular to the midsagittal plane passing through the designated impact point. (4) Adjust the dummy so that its midsagittal plane is vertical and the rear surfaces of the thorax and...

Barbell-shaped stir bar sorptive extraction using dummy template molecularly imprinted polymer coatings for analysis of bisphenol A in water.

PubMed

Liu, Ruimei; Feng, Feng; Chen, Guolin; Liu, Zhimin; Xu, Zhigang

2016-07-01

This study reports the development of a novel dummy template molecularly imprinted polymer (MIP)-coated barbell-shaped stir bar. The MIP stir bar coatings were prepared by using 2,2-bis(4-hydroxyphenyl)butane (BPB), 4,4'-dihydroxydiphenylmethane (BPF), 4-tert-butylphenol (PTBP), and tetrabromobisphenol A (TBBA) as dummy templates using a capillary in situ polymerization method. Uniform coatings can be prepared controllably. The method is simple, easy, and reproducible. The barbell-shaped stir bar was developed by using medical silicone tubes as wheels. The wheels could be removed and reinstalled when necessary; therefore, the barbell-shaped stir bar was easy to disassemble and reassemble. The novel MIP-coated stir bar showed good selectivity for the target analyte, bisphenol A (BPA). The established method is selective and sensitive with a lower detection limit for BPA of 0.003 μg/L. The dummy template MIP-coated stir bar is suitable for trace BPA analysis in real environmental water samples without template leakage. The novel stir bar can be used at least 100 times.

Measurement of Spindle Rigidity by using a Magnet Loader

NASA Astrophysics Data System (ADS)

Yamazaki, Taku; Matsubara, Atsushi; Fujita, Tomoya; Muraki, Toshiyuki; Asano, Kohei; Kawashima, Kazuyuki

The static rigidity of a rotating spindle in the radial direction is investigated in this research. A magnetic loading device (magnet loader) has been developed for the measurement. The magnet loader, which has coils and iron cores, generates the electromagnetic force and attracts a dummy tool attached to the spindle. However, the eddy current is generated in the dummy tool with the spindle rotation and reduces the attractive force at high spindle speed. In order to understand the magnetic flux and eddy current in the dummy tool, the electromagnetic field analysis by FEM was carried out. Grooves on the attraction surface of the dummy tool were designed to cut the eddy current flow. The dimension of the groove were decided based on the FEM analysis, and the designed tool were manufactured and tested. The test result shows that the designed tool successfully reduces the eddy current and recovers the attractive force. By using the magnet loader and the grooved tool, the spindle rigidity can be measured when the spindle rotates with a speed up to 10,000 min-1.

Biomechanical analysis of occupant kinematics in rollover motor vehicle accidents: dynamic spit test.

PubMed

Sances, Anthony; Kumaresan, Srirangam; Clarke, Richard; Herbst, Brian; Meyer, Steve

2005-01-01

A better understanding of occupant kinematics in rollover accidents helps to advance biomechanical knowledge and to enhance the safety features of motor vehicles. While many rollover accident simulation studies have adopted the static approach to delineate the occupant kinematics in rollover accidents, very few studies have attempted the dynamic approach. The present work was designed to study the biomechanics of restrained occupants during rollover accidents using the steady-state dynamic spit test and to address the importance of keeping the lap belt fastened. Experimental tests were conducted using an anthropometric 50% Hybrid III dummy in a vehicle. The vehicle was rotated at 180 degrees/second and the dummy was restrained using a standard three-point restraint system. The lap belt of the dummy was fastened either by using the cinching latch plate or by locking the retractor. Three configurations of shoulder belt harness were simulated: shoulder belt loose on chest with cinch plate, shoulder belt under the left arm and shoulder belt behind the chest. In all tests, the dummy stayed within the confinement of the vehicle indicating that the securely fastened lap belt holds the dummy with dynamic movement of 3 1/2" to 4". The results show that occupant movement in rollover accidents is least affected by various shoulder harness positions with a securely fastened lap belt. The present study forms a first step in delineating the biomechanics of occupants in rollover accidents.

Modal analysis of the human neck in vivo as a criterion for crash test dummy evaluation

NASA Astrophysics Data System (ADS)

Willinger, R.; Bourdet, N.; Fischer, R.; Le Gall, F.

2005-10-01

Low speed rear impact remains an acute automative safety problem because of a lack of knowledge of the mechanical behaviour of the human neck early after impact. Poorly validated mathematical models of the human neck or crash test dummy necks make it difficult to optimize automotive seats and head rests. In this study we have constructed an experimental and theoretical modal analysis of the human head-neck system in the sagittal plane. The method has allowed us to identify the mechanical properties of the neck and to validate a mathematical model in the frequency domain. The extracted modal characteristics consist of a first natural frequency at 1.3±0.1 Hz associated with head flexion-extension motion and a second mode at 8±0.7 Hz associated with antero-posterior translation of the head, also called retraction motion. Based on this new validation parameters we have been able to compare the human and crash test dummy frequency response functions and to evaluate their biofidelity. Three head-neck systems of current test dummies dedicated for use in rear-end car crash accident investigations have been evaluated in the frequency domain. We did not consider any to be acceptable, either because of excessive rigidity of their flexion-extension mode or because they poorly reproduce the head translation mode. In addition to dummy evaluation, this study provides new insight into injury mechanisms when a given natural frequency can be linked to a specific neck deformation.

Small female head and neck interaction with a deploying side airbag.

PubMed

Duma, Stefan M; Crandall, Jeff R; Rudd, Rodney W; Kent, Richard W

2003-09-01

This paper presents dummy and cadaver experiments designed to investigate the injury potential of an out-of-position small female head and neck from a deploying side airbag. Seat-mounted, thoracic-type, side airbags were selected for this study to represent those currently available on selected luxury automobiles. A computer simulation program was used to identify the worst case loading position for the small female head and neck. Once the initial position was identified, experiments were performed with the Hybrid III 5th percentile dummy and three small female cadavers, using three different inflators. Peak head center of gravity (CG) accelerations for the dummy ranged from 71x g to 154 x g, and were greater than cadaver values, which ranged from 68 x g to 103 x g. Peak neck tension as measured at the upper load cell of the dummy increased with inflator aggressivity from 992 to 1670N. A conservative modification of the US National Highway Traffic Safety Administration's (NHTSA's) N(ij) proposed neck injury criteria, which combines neck tension and bending, was used. All values were well below the 1.0 injury threshold for the dummy and suggested a very low possibility of neck injury. In agreement with this prediction, no injuries were observed. Even in a worst case position, small females are at low risk of head or neck injuries under loading from these thoracic-type airbags; however, injury risk increases with increasing inflator aggressivity.

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

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.

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

DOEpatents

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

JACKETED FUEL ELEMENT

DOEpatents

Wigner, E.P.; Szilard, L.; Creutz, E.C.

1959-02-01

These fuel elements are comprised of a homogeneous metallic uranium body completely enclosed and sealed in an aluminum cover. The uranium body and aluminum cover are bonded together by a layer of zinc located between them. The bonding layer serves to improve transfer of heat, provides an additional protection against corrosion of the uranium by the coolant, and also localizes any possible corrosion by preventing travel of corrosive material along the surface of the fuel element.

Comprehensive Fuel Spray Modeling and Impacts on Chamber Acoustics in Combustion Dynamics Simulations

DTIC Science & Technology

2013-05-01

multiple swirler configurations and fuel injector locations at atmospheric pressure con- ditions. Both single-element and multiple-element LDI...the swirl number, Reynolds’ number and injector location in the LDI element. Besides the multi-phase flow characteristics, several experimen- tal...region downstream of the fuel injector on account of a sta- ble and compact precessing vortex core. Recent ex- periments conducted by the Purdue group have

FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

DOEpatents

Flint, O.

1961-01-10

Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

Physical properties of the human head: mass, center of gravity and moment of inertia.

PubMed

Yoganandan, Narayan; Pintar, Frank A; Zhang, Jiangyue; Baisden, Jamie L

2009-06-19

This paper presents a synthesis of biomedical investigations of the human head with specific reference to certain aspects of physical properties and development of anthropometry data, leading to the advancement of dummies used in crashworthiness research. As a significant majority of the studies have been summarized as reports, an effort has been made to chronologically review the literature with the above objectives. The first part is devoted to early studies wherein the mass, center of gravity (CG), and moment of inertia (MOI) properties are obtained from human cadaver experiments. Unembalmed and preserved whole-body and isolated head and head-neck experiments are discussed. Acknowledging that the current version of the Hybrid III dummy is the most widely used anthropomorphic test device in motor vehicle crashworthiness research for frontal impact applications for over 30 years, bases for the mass and MOI-related data used in the dummy are discussed. Since the development and federalization of the dummy in the United States, description of methods used to arrive at these properties form a part of the manuscript. Studies subsequent to the development of this dummy including those from the US Military are also discussed. As the head and neck are coupled in any impact, and increasing improvements in technology such as advanced airbags, and pre-tensioners and load limiters in manual seatbelts affect the kinetics of the head-neck complex, the manuscript underscores the need to pursue studies to precisely determine all the physical properties of the head. Because the most critical parameters (locations of CG and occipital condyles (OC), mass, and MOI) have not been determined on a specimen-by-specimen basis in any single study, it is important to gather these data in future experiments. These critical data will be of value for improving occupant safety, designing advanced restraint systems, developing second generation dummies, and assessing the injury mitigating characteristics of modern vehicle components in all impact modalities.

The efficacy and safety of Baoji Tablets for treating common cold with summer-heat and dampness syndrome: study protocol for a randomized controlled trial

PubMed Central

2013-01-01

Background Despite the high incidence and the economic impact of the common cold, there are still no effective therapeutic options available. Although traditional Chinese medicine (TCM) is widely used in China to treat the common cold, there is still a lack of high-quality clinical trials. This article sets forth the protocol for a high-quality trial of a new TCM drug, Baoji Tablets, which is designed to treat the common cold with summer-heat and dampness syndrome (CCSDS). The trial is evaluating both the efficacy and safety of Baoji Tablets. Methods/design This study is designed as a multicenter, phase II, parallel-group, double-blind, double-dummy, randomized and placebo-controlled trial. A total of 288 patients will be recruited from four centers. The new tablets group are administered Baoji Tablets 0.9 g and dummy Baoji Pills 3.7 g. The old pills group are administered dummy Baoji Tablets 0.9 g and Baoji Pills 3.7 g. The placebo control group are administered dummy Baoji Tablets 0.9 g and dummy Baoji Pills 3.7 g. All drugs are taken three times daily for 3 days. The primary outcome is the duration of all symptoms. Secondary outcomes include the duration of primary and secondary symptoms, changes in primary and secondary symptom scores and cumulative symptom score at day 4, as well as an evaluation of treatment efficacy. Discussion This is the first multicenter, double-blind, double-dummy, randomized and placebo-controlled trial designated to treat CCSDS in an adult population from China. It will establish the basis for a scientific and objective assessment of the efficacy and safety of Baoji Tablets for treating CCSDS, and provide evidence for a phase III clinical trial. Trial registration This study is registered with the Chinese Clinical Trial Registry. The registration number is ChiCTR-TRC-13003197. PMID:24359521

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

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

Sarta, Jose A.; Castiblanco, Luis A

With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer ofmore » the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.« less

Reactant gas composition for fuel cell potential control

DOEpatents

Bushnell, Calvin L.; Davis, Christopher L.

1991-01-01

A fuel cell (10) system in which a nitrogen (N.sub.2) gas is used on the anode section (11) and a nitrogen/oxygen (N.sub.2 /O.sub.2) gaseous mix is used on the cathode section (12) to maintain the cathode at an acceptable voltage potential during adverse conditions occurring particularly during off-power conditions, for example, during power plant shutdown, start-up and hot holds. During power plant shutdown, the cathode section is purged with a gaseous mixture of, for example, one-half percent (0.5%) oxygen (O.sub.2) and ninety-nine and a half percent (99.5%) nitrogen (N.sub.2) supplied from an ejector (21) bleeding in air (24/28) into a high pressure stream (27) of nitrogen (N.sub.2) as the primary or majority gas. Thereafter the fuel gas in the fuel processor (31) and the anode section (11) is purged with nitrogen gas to prevent nickel (Ni) carbonyl from forming from the shift catalyst. A switched dummy electrical load (30) is used to bring the cathode potential down rapidly during the start of the purges. The 0.5%/99.5% O.sub.2 /N.sub.2 mixture maintains the cathode potential between 0.3 and 0.7 volts, and this is sufficient to maintain the cathode potential at 0.3 volts for the case of H.sub.2 diffusing to the cathode through a 2 mil thick electrolyte filled matrix and below 0.8 volts for no diffusion at open circuit conditions. The same high pressure gas source (20) is used via a "T" juncture ("T") to purge the anode section and its associated fuel processor (31).

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

Potter, David Charles; Taylor, Craig Michael; Coons, James Elmer

The percent void of the Fort Saint Vrain (FSV) material is estimated to be 21.1% based on the volume of the gap at the top of the drums, the volume of the coolant channels in the FSV fuel element, and the volume of the fuel handling channel in the FSV fuel element.

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.

The RID2 biofidelic rear impact dummy: a pilot study using human subjects in low speed rear impact full scale crash tests.

PubMed

Croft, Arthur C; Philippens, Mathieu M G M

2007-03-01

Human subjects and the recently developed RID2 rear impact crash test dummy were exposed to a series of full scale, vehicle-to-vehicle crash tests. To evaluate the biofidelity of the RID2 anthropometric test dummy on the basis of calculated neck injury criterion (NIC) values by comparing these values to those obtained from human subjects exposed in the very same crashes. The widely used and familiar hybrid III dummy has been said to lack biofidelity in the special application of low speed rear impact crashes. Several attempts have been made to modify this dummy with only marginal success. Two completely new dummies have been developed; the BioRID and the RID2. Neither have been tested under real world crash boundary conditions in side-by-side comparisons with live human subjects. Volunteer subjects, including a 50th percentile male, a 95th percentile male, and a 50th percentile female, were placed in the driver's seat of a vehicle and subjected to a series of three low speed rear impact crashes each. The RID2 dummy, which is modeled after a 50th percentile male, was placed in the passenger seat in each case. Both subjects and dummy were fully instrumented and acceleration-time histories were recorded. From this data, velocities of the heads and torsos were determined and both were used to calculate the NIC values for both crash test subjects and the RID2. The RID2 demonstrated generally higher head accelerations and NIC values than those of the human subjects. Most of the observed variations might be explained on the basis of differing head restraint geometry, posture, and body size. The RID2 NIC values compared most favorably with those of the 50th percentile male subject. For the whole group, the correlations between RID2 and human subjects did not reach statistical significance. The small number of test subjects and crash tests limited the statistical power of this pilot study, and the correlation between the RID2 and human subject NIC values were not statistically significant. The overall qualitative performance and biofidelity of the RID2 was reasonable when compared with the male human 50th percentile subject. Its overall higher ranges of head acceleration and calculated NIC values compared to all of the human subjects were generally consistent. This condition could likely be improved by increasing the stiffness of the RID2 neck. Biofidelic validation of the RID2 will require ongoing testing using a larger number of human subjects and varying boundary conditions. The results of this pilot study, while encouraging, should be considered preliminary.

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-01-14

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-05-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

DOEpatents

Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

1963-09-01

A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

Fuel shipment experience, fuel movements from the BMI-1 transport cask

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

Bauer, Thomas L.; Krause, Michael G

1986-07-01

The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including severalmore » factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask.« less

The Finite Element Modelling and Dynamic Characteristics Analysis about One Kind of Armoured Vehicles’ Fuel Tanks

NASA Astrophysics Data System (ADS)

Gao, Yang; Ge, Zhishang; Zhai, Weihao; Tan, Shiwang; Zhang, Feng

2018-01-01

The static and dynamic characteristics of fuel tank are studied for the armoured vehicle in this paper. The CATIA software is applied to build the CAD model of the armoured vehicles’ fuel tank, and the finite element model is established in ANSYS Workbench. The finite element method is carried out to analyze the static and dynamic mechanical properties of the fuel tank, and the first six orders of mode shapes and their frequencies are also computed and given in the paper, then the stress distribution diagram and the high stress areas are obtained. The results of the research provide some references to the fuel tanks’ design improvement, and give some guidance for the installation of the fuel tanks on armoured vehicles, and help to improve the properties and the service life of this kind of armoured vehicles’ fuel tanks.

Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

NASA Astrophysics Data System (ADS)

Lewis, operating defective fuel B. J.; Thompson, W. T.; Akbari, F.; Thompson, D. M.; Thurgood, C.; Higgs, J.

2004-07-01

A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor.

49 CFR 572.186 - Abdomen assembly.

Code of Federal Regulations, 2010 CFR

2010-10-01

... part of the dummy assembly shown in drawing 175-0000 including load sensors specified in § 572.189(e... measuring sensor in the abdomen as shown in Figure U5; (5) The impactor impacts the dummy's abdomen at 4.0 m... of the forces of the three abdominal load sensors, specified in 572.189(e), shall be not less than...

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

A method of casting a protective jacket about a ura nium fuel element that will bond completely to the uranium without the use of stringers or supports that would ordinarily produce gaps in the cast metal coating and bond is presented. Preformed endcaps of alumlnum alloyed with 13% silicon are placed on the ends of the uranium fuel element. These caps will support the fuel element when placed in a mold. The mold is kept at a ing alloy but below that of uranium so the cast metal jacket will fuse with the endcaps forming a complete covering and bond to the fuel element, which would otherwise oxidize at the gaps or discontinuities lefi in the coating by previous casting methods.

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

DOEpatents

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

1957-09-10

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

NUCLEAR REACTOR

DOEpatents

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

JACKETED FISSIONABLE MEMBER

DOEpatents

Boller, E.R.; Robinson, J.W.

1960-09-13

A fuel element design for a nuclear reactor is presented. The fuel element comprises a cylindrical fuel body having a portion of smaller diameter at each end thereof with an annular flange at the extreme ends of these portions of smaller diameter. An end cap fits over the ends of the fuel body and has an internal annular groove adapted to receive the flange. The fuel body and end caps are disposed in a cup-shaped jacket, a closure disc completing the enclosure of the fuel body, and tht caps are bonded over their entire periphery to the jacket.

U-Mo Plate Blister Anneal Interim Report

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

Francine J. Rice; Daniel M. Wachs; Adam B. Robinson

2010-10-01

Blister thresholds in fuel elements have been a longstanding performance parameter for fuel elements of all types. This behavior has yet to be fully defined for the RERTR U-Mo fuel types. Blister anneal studies that began in 2007 have been expanded to include plates from more recent RERTR experiments. Preliminary data presented in this report encompasses the early generations of the U-Mo fuel systems and the most recent but still developing fuel system. Included is an overview of relevant dispersion fuel systems for the purposes of comparison.

A small, 1400 deg Kelvin, reactor for Brayton space power systems

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Petrucco, Louis Jacob (Inventor); Daum, Wolfgang (Inventor)

2005-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Brown, Dale Marius (Inventor); Petrucco, Louis Jacob (Inventor); Lovett, Jeffery Allan (Inventor); Daum, Wolfgang (Inventor); Dunki-Jacobs, Robert John (Inventor)

2003-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Brown, Dale Marius (Inventor); Petrucco, Louis Jacob (Inventor); Lovett, Jeffery Allan (Inventor); Daum, Wolfgang (Inventor); Dunki-Jacobs, Robert John (Inventor)

1999-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Local Burn-Up Effects in the NBSR Fuel Element

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

Brown N. R.; Hanson A.; Diamond, D.

2013-01-31

This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peakingmore » relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.« less

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL

DOEpatents

Draley, J.E.; Ruther, W.E.

1959-01-27

Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

Preparation of a hollow porous molecularly imprinted polymer using tetrabromobisphenol A as a dummy template and its application as SPE sorbent for determination of bisphenol A in tap water.

PubMed

Li, Jin; Zhang, Xuebin; Liu, Yuxin; Tong, Hongwu; Xu, Yeping; Liu, Shaomin

2013-12-15

In this paper, a highly selective sample cleanup procedure combing dummy molecular imprinting and solid-phase extraction (DMIP-SPE) was developed for the isolation and determination of bisphenol A (BPA) in tap water. The novel hollow porous dummy molecularly imprinted polymer (HPDMIP) was prepared adopting a sacrificial support approach, using tetrabromobisphenol A (TBBPA), whose structure was similar to that of BPA, as the dummy template and mesoporous MCM-48 nanospheres as the support. Owing to a very short distance between the binding sites and the surface, a large surface area and a good steric structure to match its imprint molecules, the maximum adsorption capacities (Qmax) of the dummy-imprinted and non-imprinted sorbents for BPA were as high as 445 and 340 μmol g(-1), respectively, and the adsorption reached about 73% of Qmax in 10 min. Meanwhile, a method was developed for the determination of BPA using HPDMIP as a solid-phase extraction enrichment sorbent coupled with HPLC. Under the optimum experimental conditions, HPDMIP exhibited satisfactory results in the enrichment and determination of BPA in tap water with a recovery rate of 95-105%, and relative standard deviations of below 6%, and it can achieve a limit of detection as low as 3 ng mL(-1). The developed extraction protocol eliminated the effect of template leakage on quantitative analysis and could be applied for the determination of BPA in complicated functional samples. © 2013 Elsevier B.V. All rights reserved.

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.

Non-White, No More: Effect Coding as an Alternative to Dummy Coding with Implications for Higher Education Researchers

ERIC Educational Resources Information Center

Mayhew, Matthew J.; Simonoff, Jeffrey S.

2015-01-01

The purpose of this article is to describe effect coding as an alternative quantitative practice for analyzing and interpreting categorical, race-based independent variables in higher education research. Unlike indicator (dummy) codes that imply that one group will be a reference group, effect codes use average responses as a means for…

Crash tests of four low-wing twin-engine airplanes with truss-reinforced fuselage structure

NASA Technical Reports Server (NTRS)

Williams, M. S.; Fasanella, E. L.

1982-01-01

Four six-place, low-wing, twin-engine, general aviation airplane test specimens were crash tested under controlled free flight conditions. All airplanes were impacted on a concrete test surface at a nomial flight path velocity of 27 m/sec. Two tests were conducted at a -15 deg flight path angle (0 deg pitch angle and 15 deg pitch angle), and two were conducted at a -30 deg flight path angle (-30 deg pitch angle). The average acceleration time histories (crash pulses) in the cabin area for each principal direction were calculated for each crash test. In addition, the peak floor accelerations were calculated for each test as a function of aircraft fuselage longitudinal station number. Anthropomorphic dummy accelerations were analyzed using the dynamic response index and severity index (SI) models. Parameters affecting the dummy restraint system were studied; these parameters included the effect of no upper torso restraint, measurement of the amount of inertia-reel strap pullout before locking, measurement of dummy chest forward motion, and loads in the restraints. With the SI model, the dummies with no shoulder harness received head impacts above the concussive threshold.

Effects of vehicle front-end stiffness on rear seat dummies in NCAP and FMVSS208 tests.

PubMed

Sahraei, Elham; Digges, Kennerly; Marzougui, Dhafer

2013-01-01

This study is devoted to quantifying changes in mass and stiffness of vehicles tested by the National Highway Traffic Safety Administration (NHTSA) over the past 3 decades (model years 1982 to 2010) and understanding the effect of those changes on protection of rear seat occupants. A total of 1179 tests were used, and the changes in their mass and stiffness versus their model year was quantified. Additionally, data from 439 dummies tested in rear seats of NHTSA's full frontal crashes were analyzed. Dummies were divided into 3 groups based on their reference injury criteria. Multiple regressions were performed with speed, stiffness, and mass as predicting variables for head, neck, and chest injury criteria. A significant increase in mass and stiffness over model year of vehicles was observed, for passenger cars as well as large platform vehicles. The result showed a significant correlation (P-value < .05) between the increase in stiffness of the vehicles and increase in head and chest injury criteria for all dummy sizes. These results explain that stiffness is a significant contributor to previously reported decreases in protection of rear seat occupants over model years of vehicles.

Optimized scheduling technique of null subcarriers for peak power control in 3GPP LTE downlink.

PubMed

Cho, Soobum; Park, Sang Kyu

2014-01-01

Orthogonal frequency division multiple access (OFDMA) is a key multiple access technique for the long term evolution (LTE) downlink. However, high peak-to-average power ratio (PAPR) can cause the degradation of power efficiency. The well-known PAPR reduction technique, dummy sequence insertion (DSI), can be a realistic solution because of its structural simplicity. However, the large usage of subcarriers for the dummy sequences may decrease the transmitted data rate in the DSI scheme. In this paper, a novel DSI scheme is applied to the LTE system. Firstly, we obtain the null subcarriers in single-input single-output (SISO) and multiple-input multiple-output (MIMO) systems, respectively; then, optimized dummy sequences are inserted into the obtained null subcarrier. Simulation results show that Walsh-Hadamard transform (WHT) sequence is the best for the dummy sequence and the ratio of 16 to 20 for the WHT and randomly generated sequences has the maximum PAPR reduction performance. The number of near optimal iteration is derived to prevent exhausted iterations. It is also shown that there is no bit error rate (BER) degradation with the proposed technique in LTE downlink system.

Optimized Scheduling Technique of Null Subcarriers for Peak Power Control in 3GPP LTE Downlink

PubMed Central

Park, Sang Kyu

2014-01-01

Orthogonal frequency division multiple access (OFDMA) is a key multiple access technique for the long term evolution (LTE) downlink. However, high peak-to-average power ratio (PAPR) can cause the degradation of power efficiency. The well-known PAPR reduction technique, dummy sequence insertion (DSI), can be a realistic solution because of its structural simplicity. However, the large usage of subcarriers for the dummy sequences may decrease the transmitted data rate in the DSI scheme. In this paper, a novel DSI scheme is applied to the LTE system. Firstly, we obtain the null subcarriers in single-input single-output (SISO) and multiple-input multiple-output (MIMO) systems, respectively; then, optimized dummy sequences are inserted into the obtained null subcarrier. Simulation results show that Walsh-Hadamard transform (WHT) sequence is the best for the dummy sequence and the ratio of 16 to 20 for the WHT and randomly generated sequences has the maximum PAPR reduction performance. The number of near optimal iteration is derived to prevent exhausted iterations. It is also shown that there is no bit error rate (BER) degradation with the proposed technique in LTE downlink system. PMID:24883376

FUEL ELEMENT

DOEpatents

Fortescue, P.; Zumwalt, L.R.

1961-11-28

A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

Transformer current sensor for superconducting magnetic coils

DOEpatents

Shen, Stewart S.; Wilson, C. Thomas

1988-01-01

A transformer current sensor having primary turns carrying a primary current for a superconducting coil and secondary turns only partially arranged within the primary turns. The secondary turns include an active winding disposed within the primary turns and a dummy winding which is not disposed in the primary turns and so does not experience a magnetic field due to a flow of current in the primary turns. The active and dummy windings are wound in opposite directions or connected in series-bucking relationship, and are exposed to the same ambient magnetic field. Voltages which might otherwise develop in the active and dummy windings due to ambient magnetic fields thus cancel out. The resultant voltage is purely indicative of the rate of change of current flowing in the primary turns.

FUEL ELEMENT CONSTRUCTION

DOEpatents

Simnad, M.T.

1961-08-15

A method of preventing diffusible and volatile fission products from diffusing through a fuel element container and contaminating reactor coolant is described. More specifically, relatively volatile and diffusible fission products either are adsorbed by or react with magnesium fluoride or difluoride to form stable, less volatile, less diffusible forms. The magnesium fluoride or difluoride is disposed anywhere inwardly from the outer surface of the fuel element container in order to be contacted by the fission products before they reach and contaminate the reactor coolant. (AEC)

The Guardian: The Source for Antiterrorism Information. Volume 9, Number 1, April 2007

DTIC Science & Technology

2007-04-01

the fuel in these research reactors is generally not highly radioactive . Unlike the fuel rods in a nuclear power plant, these fuel elements would...NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING ORGANIZATION NAME(S) AND...practices and lessons learned. In addition, we will include Service and issue-specific breakout sessions that will focus on critical AT program elements

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

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

2011-01-01

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

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

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

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« less

Compact Fuel Element Environment Test

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

2012-01-01

Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

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

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

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

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

Rest, J.; Hofman, G.L.

1997-06-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U{sub 3}SiAl-Al in plate, tube, and rod configurations as a function of fission density.more » Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 {mu}m). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U{sub 3}SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs.« less

IRRADIATION METHOD AND APPARATUS

DOEpatents

Cabell, C.P.

1962-12-18

A method and apparatus are described for changing fuel bodies into a process tube of a reactor. According to this method fresh fuel elements are introduced into one end of the tube forcing used fuel elements out the other end. When sufficient fuel has been discharged, a reel and tape arrangement is employed to pull the column of bodies back into the center of the tube. Due provision is made for providing shielding in the tube. (AEC)

Yttrium and rare earth stabilized fast reactor metal fuel

DOEpatents

Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

1992-01-01

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

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

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

2008-07-15

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

Axially staggered seed-blanket reactor-fuel-module construction. [LWBR

DOEpatents

Cowell, G.K.; DiGuiseppe, C.P.

1982-10-28

A heterogeneous nuclear reactor of the seed-blanket type is provided wherein the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements.

Investigations of Multiple Swirl-Venturi Fuel Injector Concepts: Recent Experimental Optical Measurement Results for 1-Point, 7-Point, and 9-Point Configurations

NASA Technical Reports Server (NTRS)

Hicks, Yolanda R.; Anderson, Robert C.; Tedder, Sarah A.; Tacina, Kathleen M.

2015-01-01

This paper presents results obtained during testing in optically-accessible, JP8-fueled, flame tube combustors using swirl-venturi lean direct injection (LDI) research hardware. The baseline LDI geometry has 9 fuel/air mixers arranged in a 3 x 3 array within a square chamber. 2-D results from this 9-element array are compared to results obtained in a cylindrical combustor using a 7-element array and a single element. In each case, the baseline element size remains the same. The effect of air swirler angle, and element arrangement on the presence of a central recirculation zone are presented. Only the highest swirl number air swirler produced a central recirculation zone for the single element swirl-venturi LDI and the 9-element LDI, but that same swirler did not produce a central recirculation zone for the 7-element LDI, possibly because of strong interactions due to element spacing within the array.

78 FR 69943 - Anthropomorphic Test Devices; Q3s 3-Year-Old Child Side Impact Test Dummy, Incorporation by...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-21

... design and by July 2007 Build Level C was released. b. Developments In 2007, the Occupant Safety Research... reference a parts list, a set of design drawings, and a ``Procedures for Assembly, Disassembly and Inspection (PADI)'' document, to ensure that all Q3s dummies are the same in their design and construction.\\2...

Aircraft Crash Survival Design Guide. Volume 2. Aircraft Crash Environment and Human Tolerance

DTIC Science & Technology

1980-01-01

anthropometry , and crash test dummies, all of which serves as background for the design information presented in the other volumes. .I / V. L...Aeromedical Institute furnished assistance in locat- ing recent information on human tolerance, anthropometry , and crash test dummies. .3 TABLE OF CONTENTS...83 6.1 INTRODUCTION . . . . . . .. ..... 83 6.2 ANTHROPOMETRY . . . . . . 83 6.2.1 Conventional Anthropometric Measurements

SU-C-BRB-06: Utilizing 3D Scanner and Printer for Dummy Eye-Shield: Artifact-Free CT Images of Tungsten Eye-Shield for Accurate Dose Calculation

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

Park, J; Lee, J; Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul

Purpose: To evaluate the effect of a tungsten eye-shield on the dose distribution of a patient. Methods: A 3D scanner was used to extract the dimension and shape of a tungsten eye-shield in the STL format. Scanned data was transferred into a 3D printer. A dummy eye shield was then produced using bio-resin (3D systems, VisiJet M3 Proplast). For a patient with mucinous carcinoma, the planning CT was obtained with the dummy eye-shield placed on the patient’s right eye. Field shaping of 6 MeV was performed using a patient-specific cerrobend block on the 15 x 15 cm{sup 2} applicator. Themore » gantry angle was 330° to cover the planning target volume near by the lens. EGS4/BEAMnrc was commissioned from our measurement data from a Varian 21EX. For the CT-based dose calculation using EGS4/DOSXYZnrc, the CT images were converted to a phantom file through the ctcreate program. The phantom file had the same resolution as the planning CT images. By assigning the CT numbers of the dummy eye-shield region to 17000, the real dose distributions below the tungsten eye-shield were calculated in EGS4/DOSXYZnrc. In the TPS, the CT number of the dummy eye-shield region was assigned to the maximum allowable CT number (3000). Results: As compared to the maximum dose, the MC dose on the right lens or below the eye shield area was less than 2%, while the corresponding RTP calculated dose was an unrealistic value of approximately 50%. Conclusion: Utilizing a 3D scanner and a 3D printer, a dummy eye-shield for electron treatment can be easily produced. The artifact-free CT images were successfully incorporated into the CT-based Monte Carlo simulations. The developed method was useful in predicting the realistic dose distributions around the lens blocked with the tungsten shield.« less

PREIRRADIATION MEASUREMENTS OF PIQUA FUEL ELEMENTS NO. P-1111, P-1113, P- 1114, AND P-1120

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

Hubbell, H.J.

1962-11-01

Results of preirradiation measurements and tests performed during the processing and assembly of the individual fuel cylinders contained in Piqua Fuel Elements No. P-1111, P-1113, P-1114, and P-1120 are presented. A description of the techniques and equipment used in obtaining the data is also included. (auth)

Aluminum hydroxide coating thickness measurements and brushing tests on K West Basin fuel elements

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

Pitner, A.L.

1998-09-11

Aluminum hydroxide coating thicknesses were measured on fuel elements stored in aluminum canisters in K West Basin using specially developed eddy current probes . The results were used to estimate coating inventories for MCO fuel,loading. Brushing tests successfully demonstrated the ability to remove the coating if deemed necessary prior to MCO loading.

Photographic combustion characterization of LOX/hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1979-01-01

Single element injectors and two fuels were tested with the aim of photographically characterizing observed combustion phenomena. The three injectors tested were the O-F-O triplet, the transverse like on like (TLOL), and the rectangular unlike doublet (RUD). The fuels tested were RP-1 and propane. The hot firings were conducted in a specifically constructed chamber fitted with quartz windows for photographically viewing the impingement spray field. All LOX/HC testing demonstrated coking with the RP-1 fuel leaving far more soot than the propane fuel. No fuel freezing or popping was experienced under the test conditions evaluated. Carbon particle emission and combustion light brilliance increased with Pc for both fuels although RP-1 was far more energetic in this respect. The RSS phenomena appear to be present in the high Pc tests as evidenced by striations in the spray pattern and by separate fuel rich and oxidizer rich areas. The RUD element was also tested as a fuel rich gas generator element by switching the propellant circuits. Excessive sooting occurred at this low mixture ratio (0.55), precluding photographic data.

NEUTRONIC REACTOR CONTROL ELEMENT

DOEpatents

Newson, H.W.

1960-09-13

A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

ELECTROLYTIC SEPARATION PROCESS AND APPARATUS

DOEpatents

McLain, M.E. Jr.; Roberts, M.W.

1962-03-01

A method is given for dissolving stainless steel-c lad fuel elements in dilute acids such as half normal sulfuric acid. The fuel element is made the anode in a Y-shaped electrolytic cell which has a flowing mercury cathode; the stainless steel elements are entrained in the mercury and stripped therefrom by a continuous process. (AEC)

Nuclear breeder reactor fuel element with silicon carbide getter

DOEpatents

Christiansen, David W.; Karnesky, Richard A.

1987-01-01

An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

DOEpatents

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

2012-01-01

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

Dynamic Response of the Hybrid III 3 Year Old Dummy Head and Neck During Side Air Bag Loading

PubMed Central

Duma, Stefan M.; Crandall, Jeff R.; Pilkey, Walter D.; Seki, Kazuhiro; Aoki, Takashi

1998-01-01

This paper presents the results from fourteen (n = 14) tests designed to evaluate the response and injury potential of a Hybrid III 3 year old dummy subject to loading by a deploying seat mounted side air bag. An instrumented Hybrid III 3 year old dummy was used for tests in two different occupant positions chosen to maximize head and neck loading. Four seat mounted thoracic side air bags were used that varied only in the level of inflator output. NHTSA’s neck injury criteria for complex loading, referred to as Nij, was modified to include moment values for both anterioposterior and lateral directions. The results of this testing indicate that side air bag loading can result in forces and moments approaching injury threshold values. While there is considerable uncertainty as to the validity of published injury criteria due to the lack of child biomechanical data, this study demonstrates the sensitivity of child response to initial position which may provide insight into placement and geometry of side airbag systems. Furthermore, the data indicates a relationship between airbag inflator properties and child dummy response for a given airbag geometry. Recently, automobile manufacturers have begun implementing side air bags as a safety feature to mitigate injuries resulting from side impact collisions. Unlike the case for the passenger side air bag, the injury potential to an out-of-position child in side airbag loading has not been presented in the literature. The purpose of this research is to evaluate the response of a Hybrid III 3 year old dummy subject to loading by a deploying side air bag.

Development of stress boundary conditions in smoothed particle hydrodynamics (SPH) for the modeling of solids deformation

NASA Astrophysics Data System (ADS)

Douillet-Grellier, Thomas; Pramanik, Ranjan; Pan, Kai; Albaiz, Abdulaziz; Jones, Bruce D.; Williams, John R.

2017-10-01

This paper develops a method for imposing stress boundary conditions in smoothed particle hydrodynamics (SPH) with and without the need for dummy particles. SPH has been used for simulating phenomena in a number of fields, such as astrophysics and fluid mechanics. More recently, the method has gained traction as a technique for simulation of deformation and fracture in solids, where the meshless property of SPH can be leveraged to represent arbitrary crack paths. Despite this interest, application of boundary conditions within the SPH framework is typically limited to imposed velocity or displacement using fictitious dummy particles to compensate for the lack of particles beyond the boundary interface. While this is enough for a large variety of problems, especially in the case of fluid flow, for problems in solid mechanics there is a clear need to impose stresses upon boundaries. In addition to this, the use of dummy particles to impose a boundary condition is not always suitable or even feasibly, especially for those problems which include internal boundaries. In order to overcome these difficulties, this paper first presents an improved method for applying stress boundary conditions in SPH with dummy particles. This is then followed by a proposal of a formulation which does not require dummy particles. These techniques are then validated against analytical solutions to two common problems in rock mechanics, the Brazilian test and the penny-shaped crack problem both in 2D and 3D. This study highlights the fact that SPH offers a good level of accuracy to solve these problems and that results are reliable. This validation work serves as a foundation for addressing more complex problems involving plasticity and fracture propagation.

Abdominal Twin Pressure Sensors for the assessment of abdominal injuries in Q dummies: in-dummy evaluation and performance in accident reconstructions.

PubMed

Beillas, Philippe; Alonzo, François; Chevalier, Marie-Christine; Lesire, Philippe; Leopold, Franck; Trosseille, Xavier; Johannsen, Heiko

2012-10-01

The Abdominal Pressure Twin Sensors (APTS) for Q3 and Q6 dummies are composed of soft polyurethane bladders filled with fluid and equipped with pressure sensors. Implanted within the abdominal insert of child dummies, they can be used to detect abdominal loading due to the belt during frontal collisions. In the present study - which is part of the EC funded CASPER project - two versions of APTS (V1 and V2) were evaluated in abdominal belt compression tests, torso flexion test (V1 only) and two series of sled tests with degraded restraint conditions. The results suggest that the two versions have similar responses, and that the pressure sensitivity to torso flexion is limited. The APTS ability to detect abdominal loading in sled tests was also confirmed, with peak pressures typically below 1 bar when the belt loaded only the pelvis and the thorax (appropriate restraint) and values above that level when the abdomen was loaded directly (inappropriate restraint). Then, accident reconstructions performed as part of CASPER and previous EC funded projects were reanalyzed. Selected data from 19 dummies (12 Q6 and 7 Q3) were used to plot injury risk curves. Maximum pressure, maximum pressure rate and their product were all found to be injury predictors. Maximum pressure levels for a 50% risk of AIS3+ were consistent with the levels separating appropriate and inappropriate restraint in the sled tests (e.g. 50% risk of AIS3+ at 1.09 bar for pressure filtered CFC180). Further work is needed to refine the scaling techniques between ages and confirm the risk curves.

The effect of systemic lipoic acid on hearing preservation after cochlear implantation via the round window approach: A guinea pig model.

PubMed

Chang, Mun Young; Gwon, Tae Mok; Lee, Ho Sun; Lee, Jun Ho; Oh, Seung Ha; Kim, Sung June; Park, Min-Hyun

2017-03-15

The present study aimed to evaluate the effects of systemic lipoic acid on hearing preservation after cochlear implantation. Twelve Dunkin-Hartley guinea pigs were randomly divided into two groups: the control group and the lipoic acid group. Animals in the lipoic acid group received lipoic acid intraperitoneally for 4 weeks. A sterilised silicone electrode-dummy was inserted through the round window to a depth of approximately 5 mm. The hearing level was measured using auditory brainstem responses (ABRs) prior to electrode-dummy insertion, and at 4 days and 1, 2, 3 and 4 weeks after electrode-dummy insertion. The threshold shift was defined as the difference between the pre-operative threshold and each of the post-operative thresholds. The cochleae were examined histologically 4 weeks after electrode-dummy insertion. Threshold shifts changed with frequency but not time. At 2kHz, ABR threshold shifts were statistically significantly lower in the lipoic acid group than the control group. At 8, 16 and 32kHz, there was no significant difference in the ABR threshold shift between the two groups. Histologic review revealed less intracochlear fibrosis along the electrode-dummy insertion site in the lipoic acid group than in the control group. The spiral ganglion cell densities of the basal, middle and apical turns were significantly higher in the lipoic acid group compared with the control group. Therefore, systemic lipoic acid administration appears to effectively preserve hearing at low frequencies in patients undergoing cochlear implantation. These effects may be attributed to the protection of spiral ganglion cells and prevention of intracochlear fibrosis. Copyright © 2017 Elsevier B.V. All rights reserved.

On the use of an Arduino-based controller to control the charging process of a wind turbine

NASA Astrophysics Data System (ADS)

Mahmuddin, Faisal; Yusran, Ahmad Muhtam; Klara, Syerly

2017-02-01

In order to avoid an excessive charging voltage which can damage power storage when converting wind energy using a turbine, it is necessary to control the charging voltage of the turbine generator. In the present study, a charging controller which uses an Arduino microcontroller, is designed. 3 (three) indicator lights are installed to indicate the battery charging process, power diversion to dummy load and battery power level. The performance of the designed controller is evaluated by simulating 3 cases. In this simulation, a battery with maximum voltage of 12.4 V is used. Case 1 is performed with input voltage equals the one set in Arduino which is 10 V. In this case, the battery is charged up to 10.8 V. In case 2, the input voltage is 13 V while the maximum voltage set in Arduino is also 13 V. In this case, the battery is charged up to maximum voltage of the battery. Moreover, the dummy load indicator is ON and charging indicator is OFF after the maximum charging voltage is reached because the electricity is flowed to the dummy load. In the final case, the input voltage is set to be 16 V while the maximum voltage set in Arduino is 13 V. In this case, the charging indicator is OFF and dummy load indicator is ON which means that the Arduino has successfully switched the power to be flowed to dummy load. From the 3 (three) cases, it can be concluded that the designed controller works perfectly to control the charging process of the wind turbine. Moreover, the charging time needed in each case can also be determined.

Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

Liquid fuel injection elements for rocket engines

NASA Technical Reports Server (NTRS)

Cox, George B., Jr. (Inventor)

1993-01-01

Thrust chambers for liquid propellant rocket engines include three principal components. One of these components is an injector which contains a plurality of injection elements to meter the flow of propellants at a predetermined rate, and fuel to oxidizer mixture ratio, to introduce the mixture into the combustion chamber, and to cause them to be atomized within the combustion chamber so that even combustion takes place. Evolving from these injectors are tube injectors. These tube injectors have injection elements for injecting the oxidizer into the combustion chamber. The oxidizer and fuel must be metered at predetermined rates and mixture ratios in order to mix them within the combustion chamber so that combustion takes place smoothly and completely. Hence tube injectors are subject to improvement. An injection element for a liquid propellant rocket engine of the bipropellant type is provided which includes tangential fuel metering orifices, and a plurality of oxidizer tube injection elements whose injection tubes are also provided with tangential oxidizer entry slots and internal reed valves.

FLUID MODERATED REACTOR

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1957-10-22

A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

Modeling power flow in the induction cavity with a two dimensional circuit simulation

NASA Astrophysics Data System (ADS)

Guo, Fan; Zou, Wenkang; Gong, Boyi; Jiang, Jihao; Chen, Lin; Wang, Meng; Xie, Weiping

2017-02-01

We have proposed a two dimensional (2D) circuit model of induction cavity. The oil elbow and azimuthal transmission line are modeled with one dimensional transmission line elements, while 2D transmission line elements are employed to represent the regions inward the azimuthal transmission line. The voltage waveforms obtained by 2D circuit simulation and transient electromagnetic simulation are compared, which shows satisfactory agreement. The influence of impedance mismatch on the power flow condition in the induction cavity is investigated with this 2D circuit model. The simulation results indicate that the peak value of load voltage approaches the maximum if the azimuthal transmission line roughly matches the pulse forming section. The amplitude of output transmission line voltage is strongly influenced by its impedance, but the peak value of load voltage is insensitive to the actual output transmission line impedance. When the load impedance raises, the voltage across the dummy load increases, and the pulse duration at the oil elbow inlet and insulator stack regions also slightly increase.

Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

NASA Technical Reports Server (NTRS)

Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

2013-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

Improved gas tagging and cover gas combination for nuclear reactor

DOEpatents

Gross, K.C.; Laug, M.T.

1983-09-26

The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

Review of Rover fuel element protective coating development at Los Alamos

NASA Technical Reports Server (NTRS)

Wallace, Terry C.

1991-01-01

The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Foote, F.G.; Jette, E.R.

1963-05-01

A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

Extraction of the gate capacitance coupling coefficient in floating gate non-volatile memories: Statistical study of the effect of mismatching between floating gate memory and reference transistor in dummy cell extraction methods

NASA Astrophysics Data System (ADS)

Rafhay, Quentin; Beug, M. Florian; Duane, Russell

2007-04-01

This paper presents an experimental comparison of dummy cell extraction methods of the gate capacitance coupling coefficient for floating gate non-volatile memory structures from different geometries and technologies. These results show the significant influence of mismatching floating gate devices and reference transistors on the extraction of the gate capacitance coupling coefficient. In addition, it demonstrates the accuracy of the new bulk bias dummy cell extraction method and the importance of the β function, introduced recently in [Duane R, Beug F, Mathewson A. Novel capacitance coupling coefficient measurement methodology for floating gate non-volatile memory devices. IEEE Electr Dev Lett 2005;26(7):507-9], to determine matching pairs of floating gate memory and reference transistor.

Effects of external radio transmitters on fish

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

Ross, M.J.; McCormick, J.H.

1981-04-01

Yellow perch (Perca flavescens) and largemouth bass (Micropterus salmoides) were studied to determine the effects of externally attached radio transmitter tags. Perch that had been tagged with dummy radio tags were more susceptible to predation and more sensitive to environmental stress than were controls. Feeding and respiration rates were similar among dummy tagged and control groups of perch over a 6-week period. The feeding rate of dummy tagged largemouth bass was lower than that of untagged fish over a 3,5-week period. On the basis of these studies, we conclude that weights of external transmitters in water should be less thanmore » 1.5% of the fish weight. Design considerations should include streamlining components and an anterior attachment wire at the extreme leading edge of an external transmitter to prevent entanglement of the tag in surrounding vegetation.« less

Method and apparatus for diagnosing breached fuel elements

DOEpatents

Gross, K.C.; Lambert, J.D.B.; Nomura, S.

1987-03-02

The invention provides an apparatus and method for diagnosing breached fuel elements in a nuclear reactor. A detection system measures the activity of isotopes from the cover gas in the reactor. A data acquisition and processing system monitors the detection system and corrects for the effects of the cover-gas clean up system on the measured activity and further calculates the derivative curve of the corrected activity as a function of time. A plotting system graphs the derivative curve, which represents the instantaneous release rate of fission gas from a breached fuel element. 8 figs.

JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

DOEpatents

Szilard, L.; Wigner, E.P.; Creutz, E.C.

1959-05-12

Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

Enrichment Zoning Options for the Small Nuclear Rocket Engine (SNRE)

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

Bruce G. Schnitzler; Stanley K. Borowski

2010-07-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. In NASA’s recent Mars Design Reference Architecture (DRA) 5.0 study (NASA-SP-2009-566, July 2009), nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option because of its high thrust and high specific impulse (-900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. An extensive nuclear thermal rocket technology development effortmore » was conducted from 1955-1973 under the Rover/NERVA Program. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art design incorporating lessons learned from the very successful technology development program. Past activities at the NASA Glenn Research Center have included development of highly detailed MCNP Monte Carlo transport models of the SNRE and other small engine designs. Preliminary core configurations typically employ fuel elements with fixed fuel composition and fissile material enrichment. Uniform fuel loadings result in undesirable radial power and temperature profiles in the engines. Engine performance can be improved by some combination of propellant flow control at the fuel element level and by varying the fuel composition. Enrichment zoning at the fuel element level with lower enrichments in the higher power elements at the core center and on the core periphery is particularly effective. Power flattening by enrichment zoning typically results in more uniform propellant exit temperatures and improved engine performance. For the SNRE, element enrichment zoning provided very flat radial power profiles with 551 of the 564 fuel elements within 1% of the average element power. Results for this and alternate enrichment zoning options for the SNRE are compared.« less

FUEL ELEMENT CONSTRUCTION

DOEpatents

Zumwalt, L.R.

1961-08-01

Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

DOEpatents

Brooks, H.

1960-04-26

A description is given for a fuel element comprising a body of uranium metal or an uranium compound dispersed in a matrix material made from magnesium, calcium, or barium and a stainless steel jacket enclosing the body.

GAUSSIAN BEAM LASER RESONATOR PROGRAM

NASA Technical Reports Server (NTRS)

Cross, P. L.

1994-01-01

In designing a laser cavity, the laser engineer is frequently concerned with more than the stability of the resonator. Other considerations include the size of the beam at various optical surfaces within the resonator or the performance of intracavity line-narrowing or other optical elements. Laser resonators obey the laws of Gaussian beam propagation, not geometric optics. The Gaussian Beam Laser Resonator Program models laser resonators using Gaussian ray trace techniques. It can be used to determine the propagation of radiation through laser resonators. The algorithm used in the Gaussian Beam Resonator program has three major components. First, the ray transfer matrix for the laser resonator must be calculated. Next calculations of the initial beam parameters, specifically, the beam stability, the beam waist size and location for the resonator input element, and the wavefront curvature and beam radius at the input surface to the first resonator element are performed. Finally the propagation of the beam through the optical elements is computed. The optical elements can be modeled as parallel plates, lenses, mirrors, dummy surfaces, or Gradient Index (GRIN) lenses. A Gradient Index lens is a good approximation of a laser rod operating under a thermal load. The optical system may contain up to 50 elements. In addition to the internal beam elements the optical system may contain elements external to the resonator. The Gaussian Beam Resonator program was written in Microsoft FORTRAN (Version 4.01). It was developed for the IBM PS/2 80-071 microcomputer and has been implemented on an IBM PC compatible under MS DOS 3.21. The program was developed in 1988 and requires approximately 95K bytes to operate.

Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

Ion chromatographic determination of sulfur in fuels

NASA Technical Reports Server (NTRS)

Mizisin, C. S.; Kuivinen, D. E.; Otterson, D. A.

1978-01-01

The sulfur content of fuels was determined using an ion chromatograph to measure the sulfate produced by a modified Parr bomb oxidation. Standard Reference Materials from the National Bureau of Standards, of approximately 0.2 + or - 0.004% sulfur, were analyzed resulting in a standard deviation no greater than 0.008. The ion chromatographic method can be applied to conventional fuels as well as shale-oil derived fuels. Other acid forming elements, such as fluorine, chlorine and nitrogen could be determined at the same time, provided that these elements have reached a suitable ionic state during the oxidation of the fuel.

DART model for irradiation-induced swelling of uranium silicide dispersion fuel elements

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

Rest, J.; Hofman, G.L.

1999-04-01

Models for the interaction of uranium silicide dispersion fuels with an aluminum matrix, for the resultant reaction product swelling, and for the calculation of the stress gradient within the fuel particles are described within the context of DART fission-gas-induced swelling models. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by comparing DART calculations with irradiation data for the swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al in variously designed dispersion fuel elements.

Effect of strained Ge-based NMOSFETs with Ge0.93Si0.07 stressors on device layout

NASA Astrophysics Data System (ADS)

Hsu, Hung-Wen; Lee, Chang-Chun

2017-12-01

This research proposes a germanium (Ge)-based n-channel MOSFET with Ge0.93Si0.07 S/D stressor. A simulation technique is utilized to understand the layout effect of shallow trench isolation (STI) length, gate width, dummy active of diffusion (OD) length, and extended poly width on stress distribution in a channel region. Stress distribution in a channel region was simulated by ANSYS software based on finite element analysis. Furthermore, carrier mobility gain was evaluated by a second-order piezoresistance model. The piezoresistance coefficient of Ge nMOSFET varies from that of Si nMOSFET. The piezoresistance coefficient shows that longitudinal and transverse stresses are the dominant factors affecting the change in electron mobility in the channel region. For Ge-based nMOSFET, longitudinal stress tends to be tensile, whereas transverse stress tends to be compressive. Stress along channel length becomes more tensile when STI length decreases. By contrast, stress along the channel width becomes more compressive when gate width or extended poly width decreases. Electron mobility in Ge-based nMOSFET could be enhanced under the aforementioned conditions. The enhanced electron mobility becomes more significant as the device combines with a contact etching stop layer stressor. Moreover, the mobility can be improved by changing the STI length, gate width, dummy OD length, or extended poly width. This investigation systematically analyzed the relationship between layout factor and stress distribution.

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

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searchedmore » and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.« less

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Methods for making a porous nuclear fuel element

DOEpatents

Youchison, Dennis L; Williams, Brian E; Benander, Robert E

2014-12-30

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

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

A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

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

Senor, David J.; Painter, Chad L.; Geelhood, Ken J.

2007-12-01

Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling,more » core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.« less

Apollo 12 Mission image - Alan Bean unloads ALSEP RTG fuel element

NASA Image and Video Library

1969-11-19

AS12-46-6790 (19 Nov. 1969) --- Astronaut Alan L. Bean, lunar module pilot, is photographed at quadrant II of the Lunar Module (LM) during the first Apollo 12 extravehicular activity (EVA) on the moon. This picture was taken by astronaut Charles Conrad Jr., commander. Here, Bean is using a fuel transfer tool to remove the fuel element from the fuel cask mounted on the LM's descent stage. The fuel element was then placed in the Radioisotope Thermoelectric Generator (RTG), the power source for the Apollo Lunar Surface Experiments Package (ALSEP) which was deployed on the moon by the two astronauts. The RTG is next to Bean's right leg. While astronauts Conrad and Bean descended in the LM "Intrepid" to explore the Ocean of Storms region of the moon, astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) "Yankee Clipper" in lunar orbit.

Reliability analysis of dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

2008-03-01

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

MTR MAIN FLOOR. MEN DEMONSTRATE INSERTION OF DUMMY PLUG INTO ...

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

MTR MAIN FLOOR. MEN DEMONSTRATE INSERTION OF DUMMY PLUG INTO AN MTR BEAM HOLE. ONE MAN CHECKS RADIATION LEVEL AT THE END OF THE UNIVERSAL COFFIN, WHILE ANOTHER USES TOOL TO INSERT PLUG INTO HOLE THROUGH COFFIN. MEN WEAR "ANTI-C" (ANTI-CONTAMINATION) CLOTHING. INL NEGATIVE NO. 6198. R.G. Larsen, Photographer, 6/27/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Effect of occlusal interference on habitual activity of human masseter.

PubMed

Michelotti, A; Farella, M; Gallo, L M; Veltri, A; Palla, S; Martina, R

2005-07-01

It has been suggested that occlusal interference may increase habitual activity in the jaw muscles and may lead to temporomandibular disorders (TMD). We tested these hypotheses by means of a double-blind randomized crossover experiment carried out on 11 young healthy females. Strips of gold foil were glued either on a selected occlusal contact area (active interference) or on the vestibular surface of the same tooth (dummy interference) and left for 8 days each. Electromyographic masseter activity was recorded in the natural environment by portable recorders under interference-free, dummy-interference, and active-interference conditions. The active occlusal interference caused a significant reduction in the number of activity periods per hour and in their mean amplitude. The EMG activity did not change significantly during the dummy-interference condition. None of the subjects developed signs and/or symptoms of TMD throughout the whole study, and most of them adapted fairly well to the occlusal disturbance.

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.

Combined catalysts for the combustion of fuel in gas turbines

DOEpatents

Anoshkina, Elvira V.; Laster, Walter R.

2012-11-13

A catalytic oxidation module for a catalytic combustor of a gas turbine engine is provided. The catalytic oxidation module comprises a plurality of spaced apart catalytic elements for receiving a fuel-air mixture over a surface of the catalytic elements. The plurality of catalytic elements includes at least one primary catalytic element comprising a monometallic catalyst and secondary catalytic elements adjacent the primary catalytic element comprising a multi-component catalyst. Ignition of the monometallic catalyst of the primary catalytic element is effective to rapidly increase a temperature within the catalytic oxidation module to a degree sufficient to ignite the multi-component catalyst.

THE FUEL ELEMENT GRAPHITE. Project DRAGON.

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

Graham, L.W.; Price, M.S.T.

1963-01-15

The main requirements of a fuel element graphite for reactors based on the Dragon concept are low transmission coefficient for fission products, dimensional stability under service conditions, high strength, high thermal conductivity, high purity, and high resistance to oxidation. Since conclusions reached in early 1960, a considerable amount of information has accumulated concerning the likely behaviour of graphites in high temperature reactor systems, particularly data on dimensional stability under irradiation. The influence of this new knowledge on the development of fuel element graphite with the Dragon Project is discussed in detail in the final section of this paper.

Low exchange element for nuclear reactor

DOEpatents

Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

1985-01-01

A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit corrosion of certain potential NTR fuel forms. Additional diagnostic upgrades included in the present NTREES set up include the addition of a gamma ray spectrometer located near the vent filter to detect uranium fuel particles exiting the fuel element in the propellant exhaust stream to provide additional information any material loss occurring during testing. Other aspects of the upgrade included reworking NTREES to reduce the operational complexity of the system despite the increased complexity of the induction heating system. To this end, many of the controls were consolidated on fewer panels. As part of this upgrade activity, the Safety Assessment (SA) and the Standard Operating Procedures (SOPs) for NTREES were extensively rewritten. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can be accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements.

NUCLEAR REACTOR CORE

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

Bell, F.R.

1963-02-01

A nuclear reactor core composed of a number of identical elements of solid moderator material fitted together was designed. Each moderator element is apertured to provide channels for fuel and coolant. The elements have an external shape which permits them to be stacked in layers with similar elements, with the surfaces of adjacent elements fitting and in contact with each other. The cross section of the element is of a general hexagonal shape with identations and protrusions, so that the elements can be fitted together. The described core should not be liable to fracture under transverse loading. Specific arrangements ofmore » moderator elements and fuel and coolant apertures are described. (M.P.G.)« less

Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

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

Spencer, Benjamin W.; Jiang, Wen; Dolbow, John E.

As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of themore » cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply XFEM to model stress concentrations induced by fuel fractures at the fuel/cladding interface during pellet cladding mechanical interaction (PCMI). This is accomplished by enhancing the thermal and mechanical contact enforcement algorithms employed by BISON to permit their use in conjunction with XFEM. The results from this methodology are demonstrated to be equivalent to those from using meshed discrete cracks. While the results of the two methods are equivalent for the case of a stationary crack, it is demonstrated that XFEM provides the additional flexibility of allowing arbitrary crack initiation and propagation during the analysis, and minimizes model setup effort for cases with stationary cracks.« less

Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature.

PubMed

Saqib, Naeem; Bäckström, Mattias

2014-12-01

Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine content have significant effects on partitioning characteristics by increasing the formation and vaporization of highly volatile metal chlorides. Zinc and cadmium concentrations in fly ash increase with the incineration temperature. Copyright © 2014 Elsevier Ltd. All rights reserved.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-01

A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes thses overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

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

Rowsell, David Leon

This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

Demonstration of Subscale Cermet Fuel Specimen Fabrication Approach Using Spark Plasma Sintering and Diffusion Bonding

NASA Technical Reports Server (NTRS)

Barnes, Marvin W.; Tucker, Dennis S.; Benensky, Kelsa M.

2018-01-01

Nuclear thermal propulsion (NTP) has the potential to expand the limits of human space exploration by enabling crewed missions to Mars and beyond. The viability of NTP hinges on the development of a robust nuclear fuel material that can perform in the harsh operating environment (> or = 2500K, reactive hydrogen) of a nuclear thermal rocket (NTR) engine. Efforts are ongoing to develop fuel material and to assemble fuel elements that will be stable during the service life of an NTR. Ceramic-metal (cermet) fuels are being actively pursued by NASA Marshall Space Flight Center (MSFC) due to their demonstrated high-temperature stability and hydrogen compatibility. Building on past cermet fuel development research, experiments were conducted to investigate a modern fabrication approach for cermet fuel elements. The experiments used consolidated tungsten (W)-60vol%zirconia (ZrO2) compacts that were formed via spark plasma sintering (SPS). The consolidated compacts were stacked and diffusion bonded to assess the integrity of the bond lines and internal cooling channel cladding. The assessment included hot hydrogen testing of the manufactured surrogate fuel and pure W for 45 minutes at 2500 K in the compact fuel element environmental test (CFEET) system. Performance of bonded W-ZrO2 rods was compared to bonded pure W rods to access bond line integrity and composite stability. Bonded surrogate fuels retained structural integrity throughout testing and incurred minimal mass loss.

The manufacture of LEU fuel elements at Dounreay

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

Gibson, J.

1997-08-01

Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

Corrosion protected, multi-layer fuel cell interface

DOEpatents

Feigenbaum, Haim; Pudick, Sheldon; Wang, Chiu L.

1986-01-01

An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. The multi-layer configuration for the interface comprises a non-cupreous metal-coated metallic element to which is film-bonded a conductive layer by hot pressing a resin therebetween. The multi-layer arrangement provides bridging electrical contact.

A numerical investigation of the influence of radiation and moisture content on pyrolysis and ignition of a leaf-like fuel element

Treesearch

B.L. Yashwanth; B. Shotorban; S. Mahalingam; C.W. Lautenberger; David Weise

2016-01-01

The effects of thermal radiation and moisture content on the pyrolysis and gas phase ignition of a solid fuel element containing high moisture content were investigated using the coupled Gpyro3D/FDS models. The solid fuel has dimensions of a typical Arctostaphylos glandulosa leaf which is modeled as thin cellulose subjected to radiative heating on...

AST Critical Propulsion and Noise Reduction Technologies for Future Commercial Subsonic Engines Area of Interest 1.0: Reliable and Affordable Control Systems

NASA Technical Reports Server (NTRS)

Myers, William; Winter, Steve

2006-01-01

The General Electric Reliable and Affordable Controls effort under the NASA Advanced Subsonic Technology (AST) Program has designed, fabricated, and tested advanced controls hardware and software to reduce emissions and improve engine safety and reliability. The original effort consisted of four elements: 1) a Hydraulic Multiplexer; 2) Active Combustor Control; 3) a Variable Displacement Vane Pump (VDVP); and 4) Intelligent Engine Control. The VDVP and Intelligent Engine Control elements were cancelled due to funding constraints and are reported here only to the state they progressed. The Hydraulic Multiplexing element developed and tested a prototype which improves reliability by combining the functionality of up to 16 solenoids and servo-valves into one component with a single electrically powered force motor. The Active Combustor Control element developed intelligent staging and control strategies for low emission combustors. This included development and tests of a Controlled Pressure Fuel Nozzle for fuel sequencing, a Fuel Multiplexer for individual fuel cup metering, and model-based control logic. Both the Hydraulic Multiplexer and Controlled Pressure Fuel Nozzle system were cleared for engine test. The Fuel Multiplexer was cleared for combustor rig test which must be followed by an engine test to achieve full maturation.

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

Welding of unique and advanced alloys for space and high-temperature applications: welding and weldability of iridium and platinum alloys

DOE PAGES

David, Stan A.; Miller, Roger G.; Feng, Zhili

2016-08-31

Advances have been made in developing alloys for space power systems for spacecraft that travel long distances to various planets. The spacecraft are powered by radioisotope thermoelectric generators (RTGs) and the fuel element in RTGs is plutonia. For safety and containment of the radioactive fuel element, the heat source is encapsulated in iridium or platinum alloys. Ir and Pt alloys are the alloys of choice for encapsulating radioisotope fuel pellets. Ir and Pt alloys were chosen because of their high-temperature properties and compatibility with the oxide fuel element and the graphite impact shells. This review addresses the alloy design andmore » welding and weldability of Ir and Pt alloys for use in RTGs.« less

Welding of unique and advanced alloys for space and high-temperature applications: welding and weldability of iridium and platinum alloys

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

David, Stan A.; Miller, Roger G.; Feng, Zhili

Advances have been made in developing alloys for space power systems for spacecraft that travel long distances to various planets. The spacecraft are powered by radioisotope thermoelectric generators (RTGs) and the fuel element in RTGs is plutonia. For safety and containment of the radioactive fuel element, the heat source is encapsulated in iridium or platinum alloys. Ir and Pt alloys are the alloys of choice for encapsulating radioisotope fuel pellets. Ir and Pt alloys were chosen because of their high-temperature properties and compatibility with the oxide fuel element and the graphite impact shells. This review addresses the alloy design andmore » welding and weldability of Ir and Pt alloys for use in RTGs.« less

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

2010-02-23

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

HEAVY WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Szilard, L.

1958-04-29

A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

Navy Electroplating Pollution Control Technology Assessment Manual.

DTIC Science & Technology

1984-02-01

quality. Dummying of chromium baths is used in the special case where high cathode-to-anode 5ea ratio has resulted in build up of trivalent chromium (Cr...Dummying with a high anode -to-cat hode area ratio can be 6used to reoxidize the trivalent to hexavalent chromium (Cr ).Proper scheduling of work can...unit processes: * Chromium reduction (if needed) of segregated chromium waste streams to reduce the chromium from its hexavalent form to the trivalent

Synthesis of lab-in-a-pipette-tip extraction using hydrophilic nano-sized dummy molecularly imprinted polymer for purification and analysis of prednisolone.

PubMed

Arabi, Maryam; Ghaedi, Mehrorang; Ostovan, Abbas; Wang, Shaobin

2016-10-15

A novel pipette-tip based on nano-sized dummy molecularly imprinted polymer (PT-DMIP) assisted by ultrasonication for the effective enrichment and analysis of prednisolone from urine samples was developed. The PT-DMIP cartridge was prepared by packing the dummy molecularly imprinted polymer at the tip of the micropipette. The polymerization used betamethasone (BM) as the dummy template, 3-aminopropyltrimethoxysilane (APTMS) as the functionalized monomer, tetraethyl orthosilicate (TEOS) as the cross-linker and aluminum ion (Al(3+)) as a dopant to produce Lewis acid sites in the silica matrix for metal coordinative interactions with the analyte. Compared to conventional solid phase extraction (SPE), the PT-DMIP is cost-effective, fast, and easy to handle, while the system is very approachable and reduces the consumption of toxic organic solvent. HPLC-UV analysis revealed successful applicability of the sorbent for highly efficient extraction of perdnisolone from urine matrices. The extraction recovery was investigated and optimum conditions were obtained using central composite design. Good linearity for prednisolone in the range of 0.22-220μgL(-1) with regression coefficients of 0.99 reveals high applicability of the method for trace analysis. Under the optimized conditions, the recoveries are 89.0-96.1 with relative standard deviations (RSD) of less than 9.0%. Copyright © 2016 Elsevier Inc. All rights reserved.

Process-based Assignment-Setting Change for Support of Overcoming Bottlenecks in Learning by Problem-Posing in Arithmetic Word Problems

NASA Astrophysics Data System (ADS)

Supianto, A. A.; Hayashi, Y.; Hirashima, T.

2017-02-01

Problem-posing is well known as an effective activity to learn problem-solving methods. Monsakun is an interactive problem-posing learning environment to facilitate arithmetic word problems learning for one operation of addition and subtraction. The characteristic of Monsakun is problem-posing as sentence-integration that lets learners make a problem of three sentences. Monsakun provides learners with five or six sentences including dummies, which are designed through careful considerations by an expert teacher as a meaningful distraction to the learners in order to learn the structure of arithmetic word problems. The results of the practical use of Monsakun in elementary schools show that many learners have difficulties in arranging the proper answer at the high level of assignments. The analysis of the problem-posing process of such learners found that their misconception of arithmetic word problems causes impasses in their thinking and mislead them to use dummies. This study proposes a method of changing assignments as a support for overcoming bottlenecks of thinking. In Monsakun, the bottlenecks are often detected as a frequently repeated use of a specific dummy. If such dummy can be detected, it is the key factor to support learners to overcome their difficulty. This paper discusses how to detect the bottlenecks and to realize such support in learning by problem-posing.

The Nuclear Cryogenic Propulsion Stage

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Broadway, Jeramie W.; Gerrish, Harold P.; Belvin, Anthony D.; Borowski, Stanley K.; Scott, John H.

2014-01-01

Nuclear Thermal Propulsion (NTP) development efforts in the United States have demonstrated the technical viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes in a single burn (NRX-A6 test). Results from Project Rover indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. Excellent results were also obtained by the former Soviet Union. Although historical programs had promising results, many factors would affect the development of a 21st century nuclear thermal rocket (NTR). Test facilities built in the US during Project Rover no longer exist. However, advances in analytical techniques, the ability to utilize or adapt existing facilities and infrastructure, and the ability to develop a limited number of new test facilities may enable affordable development, qualification, and utilization of a Nuclear Cryogenic Propulsion Stage (NCPS). Bead-loaded graphite fuel was utilized throughout the Rover/NERVA program, and coated graphite composite fuel (tested in the Nuclear Furnace) and cermet fuel both show potential for even higher performance than that demonstrated in the Rover/NERVA engine tests.. NASA's NCPS project was initiated in October, 2011, with the goal of assessing the affordability and viability of an NCPS. FY 2014 activities are focused on fabrication and test (non-nuclear) of both coated graphite composite fuel elements and cermet fuel elements. Additional activities include developing a pre-conceptual design of the NCPS stage and evaluating affordable strategies for NCPS development, qualification, and utilization. NCPS stage designs are focused on supporting human Mars missions. The NCPS is being designed to readily integrate with the Space Launch System (SLS). A wide range of strategies for enabling affordable NCPS development, qualification, and utilization should be considered. These include multiple test and demonstration strategies (both ground and in-space), multiple potential test sites, and multiple engine designs. Two potential NCPS fuels are currently under consideration - coated graphite composite fuel and tungsten cermet fuel. During 2014 a representative, partial length (approximately 16") coated graphite composite fuel element with prototypic depleted uranium loading is being fabricated at Oak Ridge National Laboratory (ORNL). In addition, a representative, partial length (approximately 16") cermet fuel element with prototypic depleted uranium loading is being fabricated at Marshall Space Flight Center (MSFC). During the development process small samples (approximately 3" length) will be tested in the Compact Fuel Element Environmental Tester (CFEET) at high temperature (approximately 2800 K) in a hydrogen environment to help ensure that basic fuel design and manufacturing process are adequate and have been performed correctly. Once designs and processes have been developed, longer fuel element segments will be fabricated and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREE) at high temperature (approximately 2800 K) and in flowing hydrogen.

Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

NASA Technical Reports Server (NTRS)

Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

1974-01-01

The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

LCRE and SNAP 50-DR-1 programs. Engineering progress report, January 1, 1963--March 31, 1963

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

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, primary coolant circuit, aaxiliary systems, fuel elements, instrumentation, materials development, and fabrication; and SNAP-50DR-1 specifications, fuel elements, pumps, steam generator, and materials development. (DCC)

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT ...

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

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT FRAME. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-4. INEL INDEX CODE NUMBER: 075 0701 60 851 151978. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

High temperature ceramic composition for hydrogen retention

DOEpatents

Webb, R.W.

1974-01-01

A ceramic coating for H retention in fuel elements is described. The coating has relatively low thermal neutron cross section, is not readily reduced by H at 1500 deg F, is adherent to the fuel element base metal, and is stable at reactor operating temperatures. (JRD)

JACKETED URANIUM SLUG

DOEpatents

Ohlinger, L.A.; Cooper, C.M.

1958-10-01

Fuel elements for nuclear reactors are described. Eacb fuel element is comprised of a solid cylindrical slug containing fissionable material enclosed within a fluid tight jacket of neutron permeable material such as aluminum. The jacket is provided with a flexible end cap and with a sealing member having a substantially fluid-tight fit within the jacket in tight abutment with the end cap and the end of the slug. A fluid passage is provided between the end of the slug and the cap whereby leakage fiuid is principally directed to the end of the slug. In this manner, any reaction between the fissionable material and fiuid which may take place occurs more rapidly at the end of the slug than along the sides between the slug and the jacket, thereby causing longitudinal expansion of the fuel element prior to radial expansion. The longitudinal expansion can be readily detected and the fuel element removed from the coolant tube before radial expansion causes it to become jammed in the tube.

Chemical Dissolution of Simulant FCA Cladding and Plates

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

Daniel, G.; Pierce, R.; O'Rourke, P.

The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO 3-KF) flowsheets ofmore » H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.« less

Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

Fuel cell generator energy dissipator

DOEpatents

Veyo, Stephen Emery; Dederer, Jeffrey Todd; Gordon, John Thomas; Shockling, Larry Anthony

2000-01-01

An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a fuel cell generator when the electrical power output of the fuel cell generator is terminated. During a generator shut down condition, electrically resistive elements are automatically connected across the fuel cell generator terminals in order to draw current, thereby depleting the fuel

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

Studies of behavior of the fuel compound based on the U-Zr micro-heterogeneous quasialloy during cyclic thermal tests

NASA Astrophysics Data System (ADS)

Zaytsev, D. A.; Repnikov, V. M.; Soldatkin, D. M.; Solntsev, V. A.

2017-11-01

This paper provides the description of temperature cycle testing of U-Zr heterogeneous fuel composition. The composition is essentially a niobium-doped zirconium matrix with metallic uranium filaments evenly distributed over the cross section. The test samples 150 mm long had been fabricated using a fiber-filament technology. The samples were essentially two-bladed spiral mandrel fuel elements parts. In the course of experiments the following temperatures were applied: 350, 675, 780 and 1140 °C with total exposure periods equal to 200, 30, 30 and 6 hours respectively. The fuel element samples underwent post-exposure material science examination including: geometry measurements, metallographic analysis, X-ray phase analysis and electron-microscopic analysis as well as micro-hardness measurement. It has been found that no significant thermal swelling of the samples occurs throughout the whole temperature range from 350 °C up to 1140 °C. The paper presents the structural changes and redistribution of the fuel component over the fuel element cross section with rising temperature.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE PAGES

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael; ...

2016-11-17

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

DOEpatents

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

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

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

SHAPED FISSIONABLE METAL BODIES

DOEpatents

Wigner, E.P.; Williamson, R.R.; Young, G.J.

1958-10-14

A technique is presented for grooving the surface of fissionable fuel elements so that expansion can take place without damage to the interior structure of the fuel element. The fissionable body tends to develop internal stressing when it is heated internally by the operation of the nuclear reactor and at the same time is subjected to surface cooling by the circulating coolant. By producing a grooved or waffle-like surface texture, the annular lines of tension stress are disrupted at equally spaced intervals by the grooves, thereby relieving the tension stresses in the outer portions of the body while also facilitating the removal of accumulated heat from the interior portion of the fuel element.

FUEL ELEMENT INTERLOCKING ARRANGEMENT

DOEpatents

Fortescue, P.; Nicoll, D.

1963-01-01

This patent relates to a system for mutually interlocking a multiplicity of elongated, parallel, coextensive, upright reactor fuel elements so as to render a laterally selfsupporting bundle, while admitting of concurrent, selective, vertical withdrawal of a sizeable number of elements without any of the remaining elements toppling, Each element is provided with a generally rectangular end cap. When a rank of caps is aligned in square contact, each free edge centrally defines an outwardly profecting dovetail, and extremitally cooperates with its adjacent cap by defining a juxtaposed half of a dovetail- receptive mortise. Successive ranks are staggered to afford mating of their dovetails and mortises. (AEC)

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.; Kopelman, B.; Hausner, H.H.

1963-07-01

A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Hauth, J.J.; Anicetti, R.J.

1962-12-01

A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

FOIL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Noland, R.A.; Walker, D.E.; Spinrad, B.I.

1963-07-16

A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

The EPA National Fuels Surveillance Network. I. Trace constituents in gasoline and commercial gasoline fuel additives.

PubMed Central

Jungers, R H; Lee, R E; von Lehmden, D J

1975-01-01

A National Fuels Surveillance Network has been established to collect gasoline and other fuels through the 10 regional offices of the Environmental Protection Agency. Physical, chemical, and trace element analytical determinations are made on the collected fuel samples to detect components which may present an air pollution hazard or poison exhaust catalytic control devices. A summary of trace elemental constituents in over 50 gasoline samples and 18 commercially marketed consumer purchased gasoline additives is presented. Quantities of Mn, Ni, Cr, Zn, Cu, Fe, Sb, B, Mg, Pb, and S were found in most regular and premium gasoline. Environmental implications of trace constituents in gasoline are discussed. PMID:1157783

Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

NASA Astrophysics Data System (ADS)

Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

2017-07-01

Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

Complex social waves of giant honeybees provoked by a dummy wasp support the special-agent hypothesis.

PubMed

Kastberger, Gerald; Weihmann, Frank; Hoetzl, Thomas

2010-03-01

The social waves in giant honeybees termed as shimmering are more complex than mexican waves. it has been demonstrated1 that shimmering is triggered by special agents at the nest surface. in this paper, we have used a nest that originated by amalgamation of two previously separated nests and stimulated waves by a dummy wasp moved on a miniature cable car. we illustrate the plausibility of the special-agent hypothesis1 also for complex shimmering processes.

Complex social waves of giant honeybees provoked by a dummy wasp support the special-agent hypothesis

PubMed Central

Weihmann, Frank; Hoetzl, Thomas

2010-01-01

The social waves in giant honeybees termed as shimmering are more complex than mexican waves. it has been demonstrated1 that shimmering is triggered by special agents at the nest surface. in this paper, we have used a nest that originated by amalgamation of two previously separated nests and stimulated waves by a dummy wasp moved on a miniature cable car. we illustrate the plausibility of the special-agent hypothesis1 also for complex shimmering processes. PMID:20585516

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

Carpenter, John H.; Belcourt, Kenneth Noel

Completion of the CASL L3 milestone THM.CFD.P6.03 provides a tabular material properties capability to the Hydra code. A tabular interpolation package used in Sandia codes was modified to support the needs of multi-phase solvers in Hydra. Use of the interface is described. The package was released to Hydra under a government use license. A dummy physics was created in Hydra to prototype use of the interpolation routines. Finally, a test using the dummy physics verifies the correct behavior of the interpolation for a test water table. 3

Presentation of a dummy representing suit for simulation of huMAN heatloss (DRESSMAN).

PubMed

Mayer, E; Schwab, R

2004-09-01

DRESSMAN designates a novel dummy for climate measurements that allows predicting the human thermal comfort experienced inside rooms (buildings, vehicles, aircraft, railway compartments etc.) on the basis of indoor climate measurements. Measurements can be listed in tabular form and can also be represented by way of color gradations in a virtual 3D human model. Optionally, visualization may be rendered during or after measurement. Due to its very quick response, DRESSMAN is particularly suited for nonstationary processes.

SODIUM DEUTERIUM REACTOR

DOEpatents

Oppenheimer, E.D.; Weisberg, R.A.

1963-02-26

This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

2011-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

2010-04-01

The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

Axially staggered seed-blanket reactor fuel module construction

DOEpatents

Cowell, Gary K.; DiGuiseppe, Carl P.

1985-01-01

A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-07-11

Nuclear reactor fuel elements of the type in which the flssionsble material is in ceramic form, such as uranium dioxide, are described. The fuel element is comprised of elongated inner and outer concentric spaced tubular members providing an annular space therebetween for receiving the fissionable material, the annular space being closed at both ends and the inner tube being open at both ends. The fuel is in the form of compressed pellets of ceramic fissionsble material having the configuration of split bushings formed with wedge surfaces and arranged in seriated inner and outer concentric groups which are urged against the respective tubes in response to relative axial movement of the pellets in the direction toward each other. The pairs of pellets are axially urged together by a resilient means also enclosed within the annulus. This arrangement-permits relative axial displacement of the pellets during use dial stresses on the inner and outer tube members and yet maintains the fuel pellets in good thermal conductive relationship therewith.

Autothermal reforming catalyst having perovskite structure

DOEpatents

Krumpel, Michael [Naperville, IL; Liu, Di-Jia [Naperville, IL

2009-03-24

The invention addressed two critical issues in fuel processing for fuel cell application, i.e. catalyst cost and operating stability. The existing state-of-the-art fuel reforming catalyst uses Rh and platinum supported over refractory oxide which add significant cost to the fuel cell system. Supported metals agglomerate under elevated temperature during reforming and decrease the catalyst activity. The catalyst is a perovskite oxide or a Ruddlesden-Popper type oxide containing rare-earth elements, catalytically active firs row transition metal elements, and stabilizing elements, such that the catalyst is a single phase in high temperature oxidizing conditions and maintains a primarily perovskite or Ruddlesden-Popper structure under high temperature reducing conditions. The catalyst can also contain alkaline earth dopants, which enhance the catalytic activity of the catalyst, but do not compromise the stability of the perovskite structure.

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

Three-dimensional heat transfer effects during the growth of LiCaAlF 6 in a modified Bridgman furnace

NASA Astrophysics Data System (ADS)

Brandon, Simon; Derby, Jeffrey J.; Atherton, L. Jeffrey; Roberts, David H.; Vital, Russel L.

1993-09-01

A novel process modification, the simultaneous growth of three cylindrical Cr:LiCaAlf 6 (Cr:LiCAF) crystals grown from a common seed in a vertical Bridgman furnace of rectangular cross section, is assessed using computational modeling. The analysis employs the FIDAP finite-element package and accounts for three-dimensional, steady-state, conductive heat transfer throughout the system. The induction heating system is rigorously simulated via solution of Maxwell's equations. The implementation of realistic thermal boundary conditions and furnace details is shown to be important. Furnace design features are assessed through calculations, and simulations indicate expected growth conditions to be favorable. In addition, the validity of using ampoules containing "dummy" charges for experimental furnace characterization measurements is examined through test computations.

Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

NASA Technical Reports Server (NTRS)

Mayo, W.; Lantz, E.

1973-01-01

A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

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

Diamond, D. J.; Varuttamaseni, A.

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

A reload and startup plan for conversion of the NIST research reactor

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

D. J. Diamond

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2017-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). Last year NTREES was successfully used to satisfy a testing milestone for the Nuclear Cryogenic Propulsion Stage (NCPS) project and met or exceeded all required objectives.

Production test IP-544-A, irradiation of 1.6% enriched thick walled single tube elements in KER-1 and 2

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

Kratzer, W.K.; Wise, M.J.

1962-12-12

The objective of this production test is to authorize the irradiation of coextruded Zr-2 jacketed thick walled 1.6% enriched tubular elements in KER loops 1 and 2 to evaluate the swelling behavior of fuel elements at high uranium temperatures Coextruded Zr-2 jacketed 1.6% enriched tubular fuel elements 1.79 inch OD, 0.97 inch ID, and 12 inches long will be irradiated KER loops 1 and 2 to exposures no greater than 2500 MWD/T.

Validation of MCNP: SPERT-D and BORAX-V fuel

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

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Validation of MCNP: SPERT-D and BORAX-V fuel

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

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D[sup 1,2] fuel elements and BORAX-V[sup 3-8] fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Redwing: A MOOSE application for coupling MPACT and BISON

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

Frederick N. Gleicher; Michael Rose; Tom Downar

Fuel performance and whole core neutron transport programs are often used to analyze fuel behavior as it is depleted in a reactor. For fuel performance programs, internal models provide the local intra-pin power density, fast neutron flux, burnup, and fission rate density, which are needed for a fuel performance analysis. The fuel performance internal models have a number of limitations. These include effects on the intra-pin power distribution by nearby assembly elements, such as water channels and control rods, and the further limitation of applicability to a specified fuel type such as low enriched UO2. In addition, whole core neutronmore » transport codes need an accurate intra-pin temperature distribution in order to calculate neutron cross sections. Fuel performance simulations are able to model the intra-pin fuel displacement as the fuel expands and densifies. These displacements must be accurately modeled in order to capture the eventual mechanical contact of the fuel and the clad; the correct radial gap width is needed for an accurate calculation of the temperature distribution of the fuel rod. Redwing is a MOOSE-based application that enables coupling between MPACT and BISON for transport and fuel performance coupling. MPACT is a 3D neutron transport and reactor core simulator based on the method of characteristics (MOC). The development of MPACT began at the University of Michigan (UM) and now is under the joint development of ORNL and UM as part of the DOE CASL Simulation Hub. MPACT is able to model the effects of local assembly elements and is able calculate intra-pin quantities such as the local power density on a volumetric mesh for any fuel type. BISON is a fuel performance application of Multi-physics Object Oriented Simulation Environment (MOOSE), which is under development at Idaho National Laboratory. BISON is able to solve the nonlinearly coupled mechanical deformation and heat transfer finite element equations that model a fuel element as it is depleted in a nuclear reactor. Redwing couples BISON and MPACT in a single application. Redwing maps and transfers the individual intra-pin quantities such as fission rate density, power density, and fast neutron flux from the MPACT volumetric mesh to the individual BISON finite element meshes. For a two-way coupling Redwing maps and transfers the individual pin temperature field and axially dependent coolant densities from the BISON mesh to the MPACT volumetric mesh. Details of the mapping are given. Redwing advances the simulation with the MPACT solution for each depletion time step and then advances the multiple BISON simulations for fuel performance calculations. Sub-cycle advancement can be applied to the individual BISON simulations and allows multiple time steps to be applied to the fuel performance simulations. Currently, only loose coupling where data from a previous time step is applied to the current time step is performed.« less

Near-infrared bulk optical properties of goat wound tissue and human serum: consequences for an implantable optical glucose sensor.

PubMed

Aernouts, Ben; Sharma, Sandeep; Gellynck, Karolien; Vlaminck, Lieven; Cornelissen, Maria; Saeys, Wouter

2016-10-01

Near-infrared (NIR) spectroscopy offers a promising technological platform for continuous glucose monitoring in the human body. Moreover, these measurements could be performed in vivo with an implantable single-chip based optical sensor. However, a thin tissue layer may grow in the optical path of the sensor. As most biological tissues are highly scattering, they only allow a small fraction of the collimated light to pass, significantly reducing the light throughput. To quantify the effect of a thin tissue layer in the optical path, the bulk optical properties of serum and tissue samples grown on implanted dummy sensors were characterized using double integrating sphere and unscattered transmittance measurements. The estimated bulk optical properties were then used to calculate the light attenuation through a thin tissue layer. The combination band of glucose was found to be the better option, relative to the first overtone band, as the absorptivity of glucose molecules is higher, while the reduction in unscattered transmittance due to tissue growth is less. Additionally, as the wound tissue was found to be highly scattering, the unscattered transmittance of the tissue layer is expected to be very low. Therefore, a sensor configuration which measures the diffuse transmittance and/or reflectance instead was recommended. (a) Dummy sensor; (b) explanted dummy sensor in tissue lump; (c) removal of dummy sensor from tissue lump; and (d) 900 µm slices of tissue lump. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

MCNP-model for the OAEP Thai Research Reactor

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

Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculationsmore » were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.« less

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, K.C.

1988-01-21

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil area of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor. 2 figs.

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, Kenny C.

1989-01-01

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

NEUTRONIC REACTOR CONSTRUCTION

DOEpatents

Vernon, H.C.; Goett, J.J.

1958-09-01

A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

TWISTED RIBBON FUEL ELEMENT

DOEpatents

Breden, C.R.; Schultz, A.B.

1961-06-01

A reactor core formed of bundles of parallel fuel elements in the form of ribbons is patented. The fuel ribbons are twisted about their axes so as to have contact with one another at regions spaced lengthwise of the ribbons and to be out of contact with one another at locations between these spaced regions. The contact between the ribbons is sufficient to allow them to be held together in a stable bundle in a containing tube without intermediate support, while permitting enough space between the ribbon for coolant flowing.

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20

A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

Chemical characterization of biomass fuel smoke particles of rural kitchens of South Asia

NASA Astrophysics Data System (ADS)

Deka, Pratibha; Hoque, Raza Rafiqul

2015-05-01

Biomass fuel smoke particles (BFSPs) of rural kitchens collected during dry and wet seasons were characterized for elements, anions and carbon. The BFSPs of kitchens using varied biomass fuel types viz. cow dung stick, mixed biomass, cow-dung stick-mixed biomass and sugarcane bagasse were chosen for the study. The BFSPs from cow dung fuel stick showed higher levels of elements, anions and particulate carbon than other BFSPs. Calcium, K, Fe and Mg were the major elements found in all BFSPs, which did not vary much between the seasons. Sulphate was found to be the dominant anion present in all BFSPs followed by Clˉ and PO43-. Seasonal variation was pronounced in the case of abundance of anions and particulate carbon. The ratio OC/EC, often used as source signature of biomass burning, was found to be within 1.89-7.41 and 1.72-6.19 during dry and wet seasons respectively.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

ON CRITICAL MASS ANALYSIS OF JRR-2

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

None

1961-01-01

The critica mass of the JRR-2 was found to be 15 fuel elements, instead of 8 as expected, when the reactor reached criticaity. The critica mass was analyzed by AMF and JAERI a few years ago, but afterwards some modifications have been made of the stucture for the reinforcement, for example, during the construction. The critical mass is recalculated perfectly and the difference bctween 15 and S fuel elements is discussed. The deviation of the critical mass is mainly caused by the effects of control rods, fuel elcments, grid-plate, etc., in the reflector; only heavy water or light water wasmore » conaidered as the reflector in the previous calculation. A simple method is used to calculate the critical mass. The effective multiplication factor for the core with 15 fuel elements is obtained about 2% higher than the experimental value. This difference is also discussed in detail. (auth)« less

Modeling and Simulation of a Nuclear Fuel Element Test Section

NASA Technical Reports Server (NTRS)

Moran, Robert P.; Emrich, William

2011-01-01

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

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT ...

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

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT FRAME AND SUPPORT PLATFORM, AND SAFETY MECHANISM ASSEMBLY (SPRING-LOADED HINGE). F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-1. INEL INDEX CODE NUMBER: 075 0701 60 851 151975. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... transuranic elements. Different technical processes can accomplish this separation. However, over the years Purex has become the most commonly used and accepted process. Purex involves the dissolution of... facilities have process functions similar to each other, including: irradiated fuel element chopping, fuel...

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie

2014-01-01

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

NTREES Testing and Operations Status

NASA Technical Reports Server (NTRS)

Emrich, Bill

2007-01-01

Nuclear Thermal Rockets or NTR's have been suggested as a propulsion system option for vehicles traveling to the moon or Mars. These engines are capable of providing high thrust at specific impulses at least twice that of today's best chemical engines. The performance constraints on these engines are mainly the result of temperature limitations on the fuel coupled with a limited ability to withstand chemical attack by the hot hydrogen propellant. To operate at maximum efficiency, fuel forms are desired which can withstand the extremely hot, hostile environment characteristic of NTR operation for at least several hours. The simulation of such an environment would require an experimental device which could simultaneously approximate the power, flow, and temperature conditions which a nuclear fuel element (or partial element) would encounter during NTR operation. Such a simulation would allow detailed studies of the fuel behavior and hydrogen flow characteristics under reactor like conditions to be performed. Currently, the construction of such a simulator has been completed at the Marshall Space Flight Center, and will be used in the future to evaluate a wide variety of fuel element designs and the materials of which they are fabricated. This present work addresses the operational status of the Nuclear Thermal Rocket Element Environmental Simulator or NTREES and some of the design considerations which were considered prior to and during its construction.

GEH-4-42, 47; Hot pressed, I and E cooled fuel element irradiation

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

Neidner, R.

1959-11-02

In our continual effort to improve the present fuel elements which are irradiated in the numerous Hanford reactors, we have made what we believe to be a significant improvement in the hot pressing process for jacketing uranium fuel slugs. We are proposing a large scale evaluation testing program in the Hanford reactors but need the vital and basic information on the operating characteristics of this type slug under known and controlled operating conditions. We, therefore, have prepared two typical fuel slugs and will want them irradiated to about 1000 MWD/T exposure (this will require about four to five total cycles).

CERAMIC FUEL ELEMENT MATERIAL FOR A NEUTRONIC REACTOR AND METHOD OF FABRICATING SAME

DOEpatents

Duckworth, W.H.

1957-12-01

This patent relates to ceramic composition, and to neutronic reactor fuel elements formed therefrom. These ceramic elements have high density and excellent strength characteristics and are formed by conventional ceramic casting and sintering at a temperature of about 2700 deg F in a nitrogen atmosphere. The composition consists of silicon carbide, silicon, uranium oxide and a very small percentage of molybdenum. Compositions containing molybdenum are markedly stronger than those lacking this ingredient.

Full-Scale Crash Test and Finite Element Simulation of a Composite Prototype Helicopter

NASA Technical Reports Server (NTRS)

Jackson, Karen E.; Fasanella, Edwin L.; Boitnott, Richard L.; Lyle, Karen H.

2003-01-01

A full-scale crash test of a prototype composite helicopter was performed at the Impact Dynamics Research Facility at NASA Langley Research Center in 1999 to obtain data for validation of a finite element crash simulation. The helicopter was the flight test article built by Sikorsky Aircraft during the Advanced Composite Airframe Program (ACAP). The composite helicopter was designed to meet the stringent Military Standard (MIL-STD-1290A) crashworthiness criteria and was outfitted with two crew and two troop seats and four anthropomorphic dummies. The test was performed at 38-ft/s vertical and 32.5-ft/s horizontal velocity onto a rigid surface. An existing modal-vibration model of the Sikorsky ACAP helicopter was converted into a model suitable for crash simulation. A two-stage modeling approach was implemented and an external user-defined subroutine was developed to represent the complex landing gear response. The crash simulation was executed with a nonlinear, explicit transient dynamic finite element code. Predictions of structural deformation and failure, the sequence of events, and the dynamic response of the airframe structure were generated and the numerical results were correlated with the experimental data to validate the simulation. The test results, the model development, and the test-analysis correlation are described.

Potential of pedestrian protection systems--a parameter study using finite element models of pedestrian dummy and generic passenger vehicles.

PubMed

Fredriksson, Rikard; Shin, Jaeho; Untaroiu, Costin D

2011-08-01

To study the potential of active, passive, and integrated (combined active and passive) safety systems in reducing pedestrian upper body loading in typical impact configurations. Finite element simulations using models of generic sedan car fronts and the Polar II pedestrian dummy were performed for 3 impact configurations at 2 impact speeds. Chest contact force, head injury criterion (HIC(15)), head angular acceleration, and the cumulative strain damage measure (CSDM(0.25)) were employed as injury parameters. Further, 3 countermeasures were modeled: an active autonomous braking system, a passive deployable countermeasure, and an integrated system combining the active and passive systems. The auto-brake system was modeled by reducing impact speed by 10 km/h (equivalent to ideal full braking over 0.3 s) and introducing a pitch of 1 degree and in-crash deceleration of 1 g. The deployable system consisted of a deployable hood, lifting 100 mm in the rear, and a lower windshield air bag. All 3 countermeasures showed benefit in a majority of impact configurations in terms of injury prevention. The auto-brake system reduced chest force in a majority of the configurations and decreased HIC(15), head angular acceleration, and CSDM in all configurations. Averaging all impact configurations, the auto-brake system showed reductions of injury predictors from 20 percent (chest force) to 82 percent (HIC). The passive deployable countermeasure reduced chest force and HIC(15) in a majority of configurations and head angular acceleration and CSDM in all configurations, although the CSDM decrease in 2 configurations was minimal. On average a reduction from 20 percent (CSDM) to 58 percent (HIC) was recorded in the passive deployable countermeasures. Finally, the integrated system evaluated in this study reduced all injury assessment parameters in all configurations compared to the reference situations. The average reductions achieved by the integrated system ranged from 56 percent (CSDM) to 85 percent (HIC). Both the active (autonomous braking) and passive deployable system studied had a potential to decrease pedestrian upper body loading. An integrated pedestrian safety system combining the active and passive systems increased the potential of the individual systems in reducing pedestrian head and chest loading.

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

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

Renfro, David G; Chandler, David; Cook, David Howard

2014-11-01

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully convertedmore » using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.« less

Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

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

Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be wellmore » outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.« less

NASA's Nuclear Thermal Propulsion Project

NASA Technical Reports Server (NTRS)

Houts, Michael; Mitchell, Sonny; Kim, Tony; Borowski, Stanley; Power, Kevin; Scott, John; Belvin, Anthony; Clement, Steven

2015-01-01

Space fission power systems can provide a power rich environment anywhere in the solar system, independent of available sunlight. Space fission propulsion offers the potential for enabling rapid, affordable access to any point in the solar system. One type of space fission propulsion is Nuclear Thermal Propulsion (NTP). NTP systems operate by using a fission reactor to heat hydrogen to very high temperature (>2500 K) and expanding the hot hydrogen through a supersonic nozzle. First generation NTP systems are designed to have an Isp of approximately 900 s. The high Isp of NTP enables rapid crew transfer to destinations such as Mars, and can also help reduce mission cost, improve logistics (fewer launches), and provide other benefits. However, for NTP systems to be utilized they must be affordable and viable to develop. NASA's Advanced Exploration Systems (AES) NTP project is a technology development project that will help assess the affordability and viability of NTP. Early work has included fabrication of representative graphite composite fuel element segments, coating of representative graphite composite fuel element segments, fabrication of representative cermet fuel element segments, and testing of fuel element segments in the Compact Fuel Element Environmental Tester (CFEET). Near-term activities will include testing approximately 16" fuel element segments in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES), and ongoing research into improving fuel microstructure and coatings. In addition to recapturing fuels technology, affordable development, qualification, and utilization strategies must be devised. Options such as using low-enriched uranium (LEU) instead of highly-enriched uranium (HEU) are being assessed, although that option requires development of a key technology before it can be applied to NTP in the thrust range of interest. Ground test facilities will be required, especially if NTP is to be used in conjunction with high value or crewed missions. There are potential options for either modifying existing facilities or constructing new ground test facilities. At least three potential options exist for reducing (or eliminating) the release of radioactivity into the environment during ground testing. These include fully containing the NTP exhaust during the ground test, scrubbing the exhaust, or utilizing an existing borehole at the Nevada National Security Site (NNSS) to filter the exhaust. Finally, the project is considering the potential for an early flight demonstration of an engine very similar to one that could be used to support human Mars or other ambitious missions. The flight demonstration could be an important step towards the eventual utilization of NTP.

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS, ...

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

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS, FUEL ELEMENT CUTTING FACILITY, AND DRY GRAPHITE STORAGE FACILITY. INL DRAWING NUMBER 200-0603-00-030-056329. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

40 CFR 80.1141 - Small refinery exemption.

Code of Federal Regulations, 2010 CFR

2010-07-01

... (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Renewable Fuel Standard § 80.1141 Small refinery exemption...)), is exempt from the renewable fuel standards of § 80.1105 and the requirements that apply to obligated... refinery application. The application must contain all of the elements required for small refinery...

An Investigation of Technologies for Hazardous Sludge Reduction at AFLC (Air Force Logistics Command) Industrial Waste Treatment Plants. Volume 2. Literature Review of Available Technologies for Treating Heavy Metal Wastewaters.

DTIC Science & Technology

1983-12-01

plate quality. Dummying of chromium baths is used in the special case where high cathode-to-anode 15ea ratio has resulted in build up of trivalent ... chromium (Cr ).Dummying with a high anode -to-cat hode area ratio can be used to reoxidize the trivalent to hexavalent chromium (Cr +6 ). Proper...the trivalent state, which then can be precipitated as chromium hydroxide by alkali neutralization " Cyanide oxidation (if needed) of segregated

NASA general aviation crashworthiness seat development

NASA Technical Reports Server (NTRS)

Fasanella, E. L.; Alfaro-Bou, E.

1979-01-01

Three load limiting seat concepts for general aviation aircraft designed to lower the deceleration of the occupant in the event of a crash were sled tested and evaluated with reference to a standard seat. Dummy pelvis accelerations were reduced up to 50 percent with one of the concepts. Computer program MSOMLA (Modified Seat Occupant Model for Light Aircraft) was used to simulate the behavior of a dummy passenger in a NASA full-scale crash test of a twin engine light aircraft. A computer graphics package MANPLOT was developed to pictorially represent the occupant and seat motion.

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.

Electric cartridge-type heater for producing a given non-uniform axial power distribution

DOEpatents

Clark, D.L.; Kress, T.S.

1975-10-14

An electric cartridge heater is provided to simulate a reactor fuel element for use in safety and thermal-hydraulic tests of model nuclear reactor systems. The electric heat-generating element of the cartridge heater consists of a specifically shaped strip of metal cut with variable width from a flat sheet of the element material. When spirally wrapped around a mandrel, the strip produces a coiled element of the desired length and diameter. The coiled element is particularly characterized by an electrical resistance that varies along its length due to variations in strip width. Thus, the cartridge heater is constructed such that it will produce a more realistic simulation of the actual nonuniform (approximately ''chopped'' cosine) power distribution of a reactor fuel element.

Modeling 3D PCMI using the Extended Finite Element Method with higher order elements

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

Jiang, W.; Spencer, Benjamin W.

2017-03-31

This report documents the recent development to enable XFEM to work with higher order elements. It also demonstrates the application of higher order (quadratic) elements to both 2D and 3D models of PCMI problems, where discrete fractures in the fuel are represented using XFEM. The modeling results demonstrate the ability of the higher order XFEM to accurately capture the effects of a crack on the response in the vicinity of the intersecting surfaces of cracked fuel and cladding, as well as represent smooth responses in the regions away from the crack.

Calculation of Free-Atom Fractions in Hydrocarbon-Fueled Rocket Engine Plume

NASA Technical Reports Server (NTRS)

Verma, Satyajit

2006-01-01

Free atom fractions (Beta) of nine elements are calculated in the exhaust plume of CH4- oxygen and RP-1-oxygen fueled rocket engines using free energy minimization method. The Chemical Equilibrium and Applications (CEA) computer program developed by the Glenn Research Center, NASA is used for this purpose. Data on variation of Beta in both fuels as a function of temperature (1600 K - 3100 K) and oxygen to fuel ratios (1.75 to 2.25 by weight) is presented in both tabular and graphical forms. Recommendation is made for the Beta value for a tenth element, Palladium. The CEA computer code was also run to compare with experimentally determined Beta values reported in literature for some of these elements. A reasonable agreement, within a factor of three, between the calculated and reported values is observed. Values reported in this work will be used as a first approximation for pilot rocket engine testing studies at the Stennis Space Center for at least six elements Al, Ca, Cr, Cu, Fe and Ni - until experimental values are generated. The current estimates will be improved when more complete thermodynamic data on the remaining four elements Ag, Co, Mn and Pd are added to the database. A critique of the CEA code is also included.