Calcaneal traction pin placement simplified: a cadaveric study.

PubMed

Kwon, John Y; Ellington, J Kent; Marsland, Daniel; Gupta, Sanjeev

2011-06-01

The tibial neurovascular bundle and sural nerve are at risk with errant pin placement during transcalcaneal pin placement. The purpose of this study was to determine a relative safe zone using a single osseous landmark to establish a technique applicable in the presence of trauma. We describe the neural anatomy anatomically and radiographically, giving surgeons a reliable and relatively safe technique for transcalcaneal pin placement. Twenty-four cadavers were dissected for the major medial neurovascular structures and the sural nerve. The closest distance from the neurovascular structures to the posterior inferior calcaneus was measured. The mean distance from the posterior inferior calcaneus to the closest major medial neurovascular structure was 3.4 cm (SD ± 0.36; range, 2.6 to 4.1 cm). The mean distance to the sural nerve was 3.4 cm (SD ± 0.54; range, 2.3 to 4.6 cm). According to the 95% confidence intervals, a relative safe zone of 3.1 cm as a radius from the posterior inferior calcaneus was determined. A relatively safe zone of 3.1 cm based on 95% confidence intervals as described as a radius from the posterior inferior calcaneus can be used for transcalcaneal pin placement in most cases without injury to the medial neurovascular bundle or sural nerve. However anatomic variation may result in the neurovascular bundle being within this zone. We describe a surgical technique for reliable placement of a transcalcaneal pin within this relative safe zone and a safe distance from the closest neurovascular structure.

A momentum source model for wire-wrapped rod bundles—Concept, validation, and application

DOE PAGES

Hu, Rui; Fanning, Thomas H.

2013-06-19

Large uncertainties still exist in the treatment of wire-spacers and drag models for momentum transfer in current lumped parameter models. Here, to improve the hydraulic modeling of wire-wrap spacers in a rod bundle, a three-dimensional momentum source model (MSM) has been developed to model the anisotropic flow without the need to resolve the geometric details of the wire-wraps. The MSM is examined for 7-pin and 37-pin bundles steady-state simulations using the commercial CFD code STAR-CCM+. The calculated steady-state inter-subchannel cross flow velocities match very well in comparisons between bare bundles with the MSM applied and the wire-wrapped bundles with explicitmore » geometry. The validity of the model is further verified by mesh and parameter sensitivity studies. Furthermore, the MSM is applied to a 61-pin EBR-II experimental subassembly for both steady state and PLOF transient simulations. Reasonably accurate predictions of temperature, pressure, and fluid flow velocities have been achieved using the MSM for both steady-state and transient conditions. Significant computing resources are saved with the MSM since it can be used on a much coarser computational mesh.« less

Comparison of COBRA III-C and SABRE-1 (wire-wrap version) computational results with steady-state data from a 19-pin internally guard heated sodium-cooled bundle with a six-channel central blockage (THORS Bundle 3C). [LMFBR

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

Dearing, J F; Nelson, W R; Rose, S D

Computational thermal-hydraulic models of a 19-pin, electrically heated, wire-wrap liquid-metal fast breeder reactor test bundle were developed using two well-known subchannel analysis codes, COBRA III-C and SABRE-1 (wire-wrap version). These two codes use similar subchannel control volumes for the finite difference conservation equations but vary markedly in solution strategy and modeling capability. In particular, the empirical wire-wrap-forced diversion crossflow models are different. Surprisingly, however, crossflow velocity predictions of the two codes are very similar. Both codes show generally good agreement with experimental temperature data from a test in which a large radial temperature gradient was imposed. Differences between data andmore » code results are probably caused by experimental pin bowing, which is presently the limiting factor in validating coded empirical models.« less

Apparatus for inspecting fuel elements

DOEpatents

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

1986-01-01

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

Apparatus for inspecting fuel elements

DOEpatents

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

1984-12-21

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

Fuel transfer system

DOEpatents

Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.

Fuel transfer system

DOEpatents

Townsend, H.E.; Barbanti, G.

1994-03-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool. 6 figures.

Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.

1973-01-01

The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.

Nondestrucive analysis of fuel pins

DOEpatents

Stepan, I.E.; Allard, N.P.; Suter, C.R.

1972-11-03

Disclosure is made of a method and a correspondingly adapted facility for the nondestructive analysis of the concentation of fuel and poison in a nuclear reactor fuel pin. The concentrations of fuel and poison in successive sections along the entire length of the fuel pin are determined by measuring the reactivity of a thermal reactor as each successive small section of the fuel pin is exposed to the neutron flux of the reactor core and comparing the measured reactivity with the reactivities measured for standard fuel pins having various known concentrations. Only a small section of the length of the fuel pin is exposed to the neutron flux at any one time while the remainder of the fuel pin is shielded from the neutron flux. In order to expose only a small section at any one time, a boron-10-lined dry traverse tube is passed through the test region within the core of a low-power thermal nuclear reactor which has a very high fuel sensitivity. A narrow window in the boron-10 lining is positioned at the core center line. The fuel pins are then systematically traversed through the tube past the narrow window such that successive small sections along the length of the fuel pin are exposed to the neutron flux which passes through the narrow window.

Pretest predictions for degraded shutdown heat-removal tests in THORS-SHRS Assembly 1. [LMFBR

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

Rose, S.D.; Carbajo, J.J.

The recent modification of the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility at ORNL will allow testing of parallel simulated fuel assemblies under natural-convection and low-flow forced-convection conditions similar to those that might occur during a partial failure of the Shutdown Heat Removal System (SHRS) of an LMFBR. An extensive test program has been prepared and testing will be started in September 1983. THORS-SHRS Assembly 1 consists of two 19-pin bundles in parallel with a third leg serving as a bypass line and containing a sodium-to-sodium intermediate heat exchanger. Testing at low powers wil help indicate the maximum amount of heat thatmore » can be removed from the reactor core during conditions of degraded shutdown heat removal. The thermal-hydraulic behavior of the test bundles will be characterized for single-phase and two-phase conditions up to dryout. The influence of interassembly flow redistribution including transients from forced- to natural-convection conditions will be investigated during testing.« less

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

A ``NEW'' Solid-Core Reactor Fuel Form that Maximizes the Performance of Nuclear Thermal and Electric Rockets

NASA Astrophysics Data System (ADS)

Rom, Frank E.; Finnegan, Patrick M.

1994-07-01

The ``NEW'' solid-core fuel form is the old Vapor Transport (VT) fuel pin investigated at NASA about 30 years ago. It is simply a tube sealed at both ends partially filled with UO2. During operation the UO2 forms an annular layer on the inside of the tube by vaporization and condensation. This form is an ideal structure for overall strength and retention of fission products. All of the structural material lies between the fuel (including fission products) and the reactor coolant. The isothermal inside fuel surface temperature that results from the vaporization and condensation of fuel during operation eliminates hotspots, significantly increasing the design fuel pin surface temperature. For NTP, W-UO2 fuel pins yield higher operating temperatures than for other fuel forms, because W has about a ten-fold lower vaporization rate compared to any other known material. The use of perigee propulsion using W-UO2 fuel pins can result in a more than ten-fold reduction in reactor power. Lower reactor power, together with zero fission product release potential, and the simplicity of fabrication of VT fuel pins should greatly simplify and reduce the cost of development of NTP. For NEP, VT fuel pins can increase fast neutron spectrum reactor life with no fission product release. Thermal spectrum NEP reactors using W184 or Mo VT fuel pins, with only small amounts of high neutron absorbing additives, offer benefits because of much lower fissionable fuel requirements. The VT fuel pin has application to commercial power reactors with similar benefits.

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

NASA Technical Reports Server (NTRS)

Thoms, K. R.

1975-01-01

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.

Automated fuel pin loading system

DOEpatents

Christiansen, David W.; Brown, William F.; Steffen, Jim M.

1985-01-01

An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inserted as a batch prior to welding of end caps by one of two disclosed welding systems.

Automated fuel pin loading system

DOEpatents

Christiansen, D.W.; Brown, W.F.; Steffen, J.M.

An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inerted as a batch prior to welding of end caps by one of two disclosed welding systems.

Polymer Surface Textured with Nanowire Bundles to Repel High-Speed Water Drops.

PubMed

Li, Y P; Li, X Y; Zhu, X P; Lei, M K; Lakhtakia, A

2018-05-22

Water drops impacting windshields of high-speed trains and aircraft as well as blades in steam turbine power generators obliquely and at high speeds are difficult to repel. Impacting drops penetrate the void regions of nanotextured and microtextured superhydrophobic coatings, with this pinning resulting in the loss of drop mobility. In order to repel high-speed water drops, we nanotextured polymer surfaces with nanowire bundles separated from their neighbors by microscale void regions, with the nanowires in a bundle separated from their neighbors by nanoscale void regions. Water drops with speeds below a critical speed rebound completely. Water drops with speeds exceeding a critical speed rebound partially, but residual droplets that begin to be pinned undergo a spontaneous dewetting process and slide off. The natural oscillations of residual droplets drive this dewetting process in the interbundle void regions, resulting in a transition from the sticky Wenzel state to the slippery Cassie state without external stimuli.

Fuel pin

DOEpatents

Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

1987-11-24

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

Fuel pin

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

1989-10-03

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

Fuel pin

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

1989-01-01

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

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

NASA Astrophysics Data System (ADS)

Katsuyama, Kozo; Nagamine, Tsuyoshi; Furuya, Hirotaka

2010-10-01

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

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

Ingham, J.G.

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

Evaluation of FFTF fuel pin design procedure vis-a-vis steady state irradiation performance in EBR II

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

Jackson, R.J.

1976-11-01

The FFTF fuel pin design analysis is shown to be conservative through comparison with pin irradiation experience in EBR-II. This comparison shows that the actual lifetimes of EBR-II fuel pins are either greater than 80,000 MWd/MTM or greater than the calculated allowable lifetimes based on thermal creep strain.

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

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

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

2012-07-01

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

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

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

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

2012-04-01

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

Analysis of the OPERA 15-pin experiment with SABRE-2P. [LMFBR

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

Rose, S.D.; Carbajo, J.J.

The OPERA (Out-of-Pile Expulsion and Reentry Apparatus) experiment simulates the initial phase of a pump coastdown without scram of a liquid-metal fast breeder reactor, specifically the Fast Flux Test Facility. The test section is a 15-pin 60/sup 0/ triangular sector designed to simulate a full-size 61-pin hexagonal bundle. A previous study indicates this to be an adequate simulation. In this paper, experimental results from the OPERA 15-pin experiment performed at ANL in 1982 are compared to analytical calculations obtained with the SABRE-2P code at ORNL.

Two-phase pressure drop reduction BWR assembly design

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

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

1992-05-12

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

Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

NASA Astrophysics Data System (ADS)

Porter, D. L.; Tsai, Hanchung

2012-08-01

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could actually help extend the life of a fuel pin in a "long pin" reactor design to a higher peak fuel burnup.

Retrievable fuel pin end member for a nuclear reactor

DOEpatents

Rosa, Jerry M.

1982-01-01

A bottom end member (17b) on a retrievable fuel pin (13b) secures the pin (13b) within a nuclear reactor (12) by engaging on a transverse attachment rail (18) with a spring clip type of action. Removal and reinstallation if facilitated as only axial movement of the fuel pin (13b) is required for either operation. A pair of resilient axially extending blades (31) are spaced apart to define a slot (24) having a seat region (34) which receives the rail (18) and having a land region (37), closer to the tips (39) of the blades (31) which is normally of less width than the rail (18). Thus an axially directed force sufficient to wedge the resilient blades (31) apart is required to emplace or release the fuel pin (13b) such force being greater than the axial forces on the fuel pins (13b) which occur during operation of the reactor (12).

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

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

Mays, Brian; Jackson, R. Brian

2017-03-08

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

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.

1973-01-01

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

Seed and blanket fuel arrangement for dual-phase nuclear reactors

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

Congdon, S.P.; Fawcett, R.M.

1992-09-22

This patent describes a fuel management method for a dual-phase nuclear reactor, it comprises: installing a fuel bundle at a first core location accessed by coolant through a relatively small aperture, each of the bundles having a predetermined group of fuel elements; operating the reactor a first time; shutting down the reactor; reinstalling the fuel bundle at a second core location accessed by coolant through a relatively large aperture; and operating the reactor a second time.

EXFILE: A program for compiling irradiation data on UN and UC fuel pins

NASA Technical Reports Server (NTRS)

Mayer, J. T.; Smith, R. L.; Weinstein, M. B.; Davison, H. W.

1973-01-01

A FORTRAN-4 computer program for handling fuel pin data is described. Its main features include standardized output, easy access for data manipulation, and tabulation of important material property data. An additional feature allows simplified preparation of input decks for a fuel swelling computer code (CYGRO-2). Data from over 300 high temperature nitride and carbide based fuel pin irradiations are listed.

Carbide fuel pin and capsule design for irradiations at thermionic temperatures

NASA Technical Reports Server (NTRS)

Siegel, B. L.; Slaby, J. G.; Mattson, W. F.; Dilanni, D. C.

1973-01-01

The design of a capsule assembly to evaluate tungsten-emitter - carbide-fuel combinations for thermionic fuel elements is presented. An inpile fuel pin evaluation program concerned with clad temperture, neutron spectrum, carbide fuel composition, fuel geometry,fuel density, and clad thickness is discussed. The capsule design was a compromise involving considerations between heat transfer, instrumentation, materials compatibility, and test location. Heat-transfer calculations were instrumental in determining the method of support of the fuel pin to minimize axial temperature variations. The capsule design was easily fabricable and utilized existing state-of-the-art experience from previous programs.

Effect of Fuel Temperature Profile on Eigenvalue Calculations

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

Greifenkamp, Tom E; Clarno, Kevin T; Gehin, Jess C

2008-01-01

Use of an average fuel temperature is a current practice when modeling fuel for eigenvalue (k-inf) calculations. This is an approximation, as it is known from Heat-transfer methods that a fuel pin having linear power q', will have a temperature that varies radially and has a maximum temperature at the center line [1]. This paper describes an investigation into the effects on k-inf and isotopic concentrations of modeling a fuel pin using a single average temperature versus a radially varying fuel temperature profile. The axial variation is not discussed in this paper. A single fuel pin was modeled having 1,more » 3, 5, 8, or 10 regions of equal volumes (areas). Fig. 1 shows a model of a 10-ring fuel pin surrounded by a gap and then cladding.« less

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.

Integral Fast Reactor fuel pin processor

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

Levinskas, D.

1993-01-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

Integral Fast Reactor fuel pin processor

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

Levinskas, D.

1993-03-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

Thermal analysis of the FSP-1 fuel pin irradiation test. [for SP-100 space power reactor

NASA Technical Reports Server (NTRS)

Lyon, William F., III

1991-01-01

Thermal analysis of a pin from the FSP-1 fuels irradiation test has been completed. The purpose of the analysis was to provide predictions of fuel pin temperatures, determine the flow regime within the lithium annulus of the test assembly, and provide a standardized model for a consistent basis of comparison between pins within the test assembly. The calculations have predicted that the pin is operating at slightly above the test design temperatures and that the flow regime within the lithium annulus is a laminar buoyancy driven flow.

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.

Fabrication of fuel pin assemblies, phase 3

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

Molybdenum-UO2 cerment irradiation at 1145 K

NASA Technical Reports Server (NTRS)

Mcdonald, G.

1971-01-01

Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

Assemblies with both target and fuel pins in an isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

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.)

A Mathematical Model for Continuous Fiber Reinforced Thermoplastic Composite in Melt Impregnation

NASA Astrophysics Data System (ADS)

Ren, Feng; Yu, Yang; Yang, Jianjun; Xin, Chunling; He, Yadong

2017-06-01

Through the combination of Reynolds equation and Darcy's law, a mathematical model was established to calculate the pressure distribution in wedge area, which contributed to the forecast effect of processing parameters on impregnation degree of the fiber bundle. The experiments were conducted to verify the capacity of the proposed model with satisfactory results, which means that the model is effective in predicting the influence of processing parameters on impregnation. From the mathematical model, it was known that the impregnation degree of the fiber bundle would be improved by increasing the processing temperature, number and radius of pins, or decreasing the pulling speed and the center distance of pins, which provided a possible solution to the difficulty of melt with high viscosity in melt impregnation and optimization of impregnation processing.

System for supporting a bundled tube fuel injector within a combustor

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

LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold

A combustor includes an end cover having an outer side and an inner side, an outer barrel having a forward end that is adjacent to the inner side of the end cover and an aft end that is axially spaced from the forward end. An inner barrel is at least partially disposed concentrically within the outer barrel and is fixedly connected to the outer barrel. A fluid conduit extends downstream from the end cover. A first bundled tube fuel injector segment is disposed concentrically within the inner barrel. The bundled tube fuel injector segment includes a fuel plenum that ismore » in fluid communication with the fluid conduit and a plurality of parallel tubes that extend axially through the fuel plenum. The bundled tube fuel injector segment is fixedly connected to the inner barrel.« less

Fuel Pin Behavior Under the Slow Power Ramp Transients in the CABRI-2 Experiments

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

Charpenel, Jean; Lemoine, Francette; Sato, Ikken

Slow ramp-type transient-overpower tests were performed within the framework of the international CABRI-2 experimental program. The implemented power transients of {approx}1% nominal power/s correspond to a control rod withdrawal-type accident in a liquid-metal-cooled fast breeder reactor (FBR). The analysis of the tests includes the information elements derived from the hodoscope signals, which were assessed quantitatively and supported by destructive and nondestructive posttest examinations. These tests, performed with fuels of various geometries, demonstrated the high margin to failure of such FBR fuel pins within the expected power level before the emergency reactor shutdown. At the same time, these tests performed withmore » high- and low-smear-density industrial pins led to clarification of the influence of pellet design on fuel pin behavior under high overpower condition. With the high-smear-density solid fuel pellet pin of high burnup level, the retained gaseous fission products played an important role in the solid fuel swelling, leading to clad deformation and failure at a maximum heating rate of 81 kW.m{sup -1}, which is much greater than the end-of-life (EOL) linear rating of the pin. With the low smear-density annular pellet pin, an important fuel swelling takes place, leading to degradation of the fuel thermal conductivity. This effect was detected at the power level around 73 kW.m{sup -1}, which is also much higher than the EOL value of the pin. Furthermore, the absence of clad deformation, and consequently of failure even at the power level going up to 134.7 kW.m{sup -1}, confirmed the very high margin to failure. In consequence, it was clarified that gaseous fission products have significant effects on failure threshold as well as on thermal performance during overpower condition, and such effects are significantly dependent on fuel design and power operation conditions.« less

Auxin efflux facilitator and auxin dynamism responsible for the gravity-regulated development of peg in cucumber seedlings

NASA Astrophysics Data System (ADS)

Takahashi, Hideyuki; Watanabe, Chiaki; Fujii, Nobuharu; Miyazawa, Yutaka

Cucumber seedlings develop a protuberance, peg, by which seed coats are pulled out just af-ter germination. The peg is usually formed on the lower side of the transition zone between hypocotyl and root of the seedlings grown in a horizontal position. Our previous spaceflight experiment showed that unilateral positioning of a peg in cucumber seedlings occurred due to its suppression on the upper side of the transition zone because seedlings grown in microgravity developed a peg on each side of the transition zone. We also showed that auxin was a major factor responsible for peg development. There was a redistribution of auxin in the gravistimu-lated transition zone, decreasing IAA level on the upper side, and IAA application induced a peg on both lower and upper sides of the transition zone. In addition, peg was released from its suppression in the seedlings treated with inhibitors of auxin efflux. Namely, two pegs devel-oped in the TIBA-treated seedlings even when they were grown in a horizontal position. These results imply that a reduction of auxin level due to its efflux is required for the suppression of peg development on the upper side of the transition zone in a horizontal position. To under-stand molecular mechanism underlying the negative control of morphogenesis by graviresponse in cucumber seedlings, we isolated cDNAs of auxin efflux facilitators, CsPINs, from cucumber and examined the expressions of their proteins, in relation to the redistribution of endogenous auxin and peg development. We isolated six cDNAs of PIN homologues CsPIN1 to CsPIN6 from cucumber. By immunohistochemical study using some of their anti-bodies, we revealed that CsPIN1 was localized in endodermis, vascular tissue and pith around the transition zone of cucumber seedlings. In cucumber seedlings grown in a vertical position with radicles pointing down, CsPIN1 in endodermal cells was mainly localized on the plasma membrane neighboring vascular bundle but not on the plasma membrane next to the cortex. This CsPIN1 localization could play a role in transporting auxin from cortex to vascular bundle. In both vascular and pith tissues, CsPIN1 was localized on the bottom plasma membrane of the cells, which could allow auxin to move toward the roots. In the seedlings grown in a horizontal position, endoder-mal cells situated above the vascular bundle localized CsPIN1 on the lower plasma membrane, whereas the polarized localization of CsPIN1 in endodermal cells situated below the vascular bundle became less clear. This differential expression of CsPIN1 in the endodermis commenced within 30 min after gravistimulation. We measured endogenous IAA contents in the transi-tion zone of the 24-hour-old seedlings. In the longitudinally halved transition zone of seedlings grown in a horizontal position, free IAA content was significantly lowered in the upper side, compared to that of the lower side or either side of the transition zone in a vertical position. When 24-hour-old seedlings grown in a vertical position were gravistimulated by reorienting them to the horizontal, free IAA in the lower side of the transition zone increased by 30 min after gravistimulation and eventually decreased to the control level by 180 min after gravistim-ulation. IAA content in the upper side of the transition zone did not change much and was comparable to that in the vertical transition zone during 180 min after gravistimulation. Thus, it appears that gravistimulation causes an immediate increase of IAA level in the lower side and its eventual decrease in the upper side of the transition zone. The gravity-induced changes in CsPIN1 localization in endodermal cells could be involved in auxin redistribution that leads to unilateral positioning of a peg in cucumber seedlings.

Theoretical analysis of swelling characteristics of cylindrical uranium dioxide fuel pins with a niobium - 1-percent-zirconium clad

NASA Technical Reports Server (NTRS)

Saltsman, J. F.

1973-01-01

The relations between clad creep strain and fuel volume swelling are shown for cylindrical UO2 fuel pins with a Nb-1Zr clad. These relations were obtained by using the computer code CYGRO-2. These clad-strain - fuel-volume-swelling relations may be used with any fuel-volume-swelling model, provided the fuel volume swelling is isotropic and independent of the clad restraints. The effects of clad temperature (over a range from 118 to 1642 K (2010 to 2960 R)), pin diameter, clad thickness and central hole size in the fuel have been investigated. In all calculations the irradiation time was 500 hours. The burnup rate was varied.

Generator module architecture for a large solid oxide fuel cell power plant

DOEpatents

Gillett, James E.; Zafred, Paolo R.; Riggle, Matthew W.; Litzinger, Kevin P.

2013-06-11

A solid oxide fuel cell module contains a plurality of integral bundle assemblies, the module containing a top portion with an inlet fuel plenum and a bottom portion receiving air inlet feed and containing a base support, the base supports dense, ceramic exhaust manifolds which are below and connect to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the fuel cells comprise a fuel cell stack bundle all surrounded within an outer module enclosure having top power leads to provide electrical output from the stack bundle, where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all 100% of the weight of the stack, and each bundle assembly has its own control for vertical and horizontal thermal expansion control.

Process-Hardened, Multi-Analyte Sensor for Characterizing Rocket Plume Constituents

NASA Technical Reports Server (NTRS)

Goswami, Kisholoy

2011-01-01

A multi-analyte sensor was developed that enables simultaneous detection of rocket engine combustion-product molecules in a launch-vehicle ground test stand. The sensor was developed using a pin-printing method by incorporating multiple sensor elements on a single chip. It demonstrated accurate and sensitive detection of analytes such as carbon dioxide, carbon monoxide, kerosene, isopropanol, and ethylene from a single measurement. The use of pin-printing technology enables high-volume fabrication of the sensor chip, which will ultimately eliminate the need for individual sensor calibration since many identical sensors are made in one batch. Tests were performed using a single-sensor chip attached to a fiber-optic bundle. The use of a fiber bundle allows placement of the opto-electronic readout device at a place remote from the test stand. The sensors are rugged for operation in harsh environments.

Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

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

Bromley, B.P.; Hyland, B.

2013-07-01

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (aboutmore » 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)« less

Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

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

Bromley, B.P.; Hyland, B.

2013-07-01

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu andmore » Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)« less

Posttest examination results of recent treat tests on metal fuel

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

Holland, J.W.; Wright, A.E.; Bauer, T.H.

A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity.more » Eutectic penetration and failure of the cladding were also examined in the failed pins.« less

The chemical state of defective uranium-plutonium oxide fuel pins irradiated in sodium cooled reactors

NASA Astrophysics Data System (ADS)

Kleykamp, H.

1997-09-01

Steady-state irradiation experiments were conducted in the sodium loop of the Siloe reactor on artificially failed mixed oxide pins that had been pre-irradiated in fast reactors up to 11.5% burnup. The formation of the predominant reaction product Na 3(U,Pu)O 4 starts on the fuel surface and is terminated when a lower O/(U + Pu) threshold of the fuel is attained. The axial extent of the reaction product depends on the size of the initial cladding defect. The occurrence of secondary cracks is possible. Na(U,Pu)O 3 forms at higher fuel temperatures. The existence of Na 3U 1- xPu xO 4 is shown in pre-irradiated blanket pins after artificial defect formation. Caesium in the oxocompounds is reduced to the metallic state and is dissolved in the coolant. Evidence of a very low chemical potential of oxygen in defective fuel pins is sustained by the occurrence of actinide-platinum metal phases formed by coupled reduction of hypostoichiometric fuel with ɛ-(Mo,Tc,Ru,Rh,Pd) precipitates. Continued operation of defective pins is not hazardous by easy precautions.

Sustained Recycle in Light Water and Sodium-Cooled Reactors

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

Steven J. Piet; Samuel E. Bays; Michael A. Pope

2010-11-01

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in freshmore » fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.« less

Irradiation of three T-111 clad uranium nitride fuel pins for 8070 hours at 990 C (1815 F)

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.; Gedeon, L.; Galbo, R. J.

1973-01-01

The design and successful operation of three tantalum alloy (Ta-8W-2Hf) clad uranium mononitride (UN) fuel pins irradiated for 8070 hr at 990 C (1815 F) is described. Two pin diameters having measured burnups of 0.47 and 0.90 uranium atom percent were tested. No clad failures or swelling was detected; however, postirradiation clad samples tested failed with 1 percent strain. The fuel density decrease was 2 percent, and the fission gas release was less than 0.05 percent. Isotropic fuel swelling, which averaged about 0.5 percent, was less than fuel pin assembly clearances. Thus the clad was not strained. Thermocouples with a modified hot zone operated at average temperatures to 1100 C (2012 F) without failure. Factors that influence the ability to maintain uniform clad temperature as well as the results of the heat transfer calculations are discussed.

Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

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

Serell, D.C.; Kaplan, S.

1980-09-01

Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

Improved perturbation method for gadolinia worth calculation

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

Chiang, R.T.; Congdon, S.P.

1986-01-01

Gadolinia is utilized in light water power reactors as burnable poison for reserving excess reactivity. Good gadolinia worth estimation is useful for evaluating fuel bundle designs, core operating strategies, and fuel cycle economics. The authors have developed an improved perturbation method based on exact perturbation theory for gadolinia worth calculations in fuel bundles. The method predicts much more accurate gadolinia worth than the first-order perturbation method (commonly used to estimate nuclide worths) for bundles containing fresh or partly burned gadolinia.

Recent developments in BWR fuel design

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

Congdon, S.P.; Noble, L.D.; Wood, J.E.

1991-11-01

Substantial increases in the cost effectiveness and performance capability of boiling water reactor (BWR) fuel designs have been implemented in the past 5 to 7 yr. This increase has been driven by (a) utility desires to lower fuel and operating costs and (b) design innovations that have lowered enrichment requirements, improved thermal-hydraulic performance, and increased discharge exposure. Higher discharge exposures reduce disposal costs for European and Asian utilities and enable US utilities to lengthen operating cycles. A typical BWR reload fuel bundle fabricated today has 25% higher {sup 235}U enrichment and a factor of 2 higher gadolinium loading than onemore » made several years ago. Today's BWR fuel bundles also contain more unheated water reduces the axial water density variation, lowers the void coefficient, and enhances the neutron efficiency of the bundle, reducing both the gadolinium poison and the enrichment requirements. In addition to these general trends, the following unique design innovations have further enhanced the fuel cost efficiency and performance characteristics of BWR fuel: ferrule spacer, part length rods, interactive channel, and bundle enhanced spectral shift. GE's fuel designs offer the flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility and fuel cycle economics.« less

Recent GE BWR fuel experience and design evolution

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

Wood, J.E.; Potts, G.A.; Proebstle, R.A.

1992-01-01

Reliable fuel operation is essential to the safe, reliable, and economic power production by today's commercial nuclear reactors. GE Nuclear Energy is committed to maximize fuel reliability through the progressive development of improved fuel design features and dedication to provide the maximum quality of the design features and dedication to provide the maximum quality of the design, fabrication, and operation of GE BWR fuel. Over the last 35 years, GE has designed, fabricated, and placed in operation over 82,000 BWR fuel bundles containing over 5 million fuel rods. This experience includes successful commercial reactor operation of fuel assemblies to greatermore » than 45000 MWd/MTU bundle average exposure. This paper reports that this extensive experience base has enabled clear identification and characterization of the active failure mechanisms. With this failure mechanism characterization, mitigating actions have been developed and implemented by GE to provide the highest reliability BWR fuel bundles possible.« less

Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

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

Bhatti, Zaki; Hyland, B.; Edwards, G.W.R.

2013-07-01

The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in themore » Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)« less

Valve for fuel pin loading system

DOEpatents

Christiansen, David W.

1985-01-01

A cyclone valve surrounds a wall opening through which cladding is projected. An axial valve inlet surrounds the cladding. Air is drawn through the inlet by a cyclone stream within the valve. An inflatable seal is included to physically engage a fuel pin subassembly during loading of fuel pellets.

Summary of LCRE fuel element design including supporting experimental data

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

None, None

Declassified 18 Sep 1973. The design basis of the LCRE fuel pin is presented. The fuel pin consists of a Cb-1 Zr alloy cladding tube 0.305 inch diameter, 0.015 inch wall thickness and 35.96 inches long. The active fuel section is 13.5 inches long, with top and bottom reflector rods each 6.9 inches long and with a 4 inch gas accumulation space at each end. The cladding is designed as a pressure vessel to contain the gases released from the fuel and end refiector materials, which results in an internal gas pressure buildup in the pins during reactor operation. (23more » referencea) (auth)« less

75 FR 7557 - Airworthiness Directives; The Boeing Company Model 767 Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-22

... shorts in many systems, including the spar fuel shut off valve, oxygen mask deployment, and burned wires... and wire bundles, causing shorts in many systems, including the spar fuel shut off valve, oxygen mask... and wire bundles, causing shorts in many systems, including the spar fuel shut off valve, oxygen mask...

Fuel bundle design for enhanced usage of plutonium fuel

DOEpatents

Reese, Anthony P.; Stachowski, Russell E.

1995-01-01

A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced.

Fuel bundle design for enhanced usage of plutonium fuel

DOEpatents

Reese, A.P.; Stachowski, R.E.

1995-08-08

A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced. 10 figs.

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Fabrication of capsule assemblies, phase 3

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Fast-spectrum space-power-reactor concepts using boron control devices

NASA Technical Reports Server (NTRS)

Mayo, W.

1973-01-01

Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.

Hybrid Gama Emission Tomography (HGET): FY16 Annual Report

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

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

2017-02-01

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

Preliminary analysis of hot spot factors in an advanced reactor for space electric power systems

NASA Technical Reports Server (NTRS)

Lustig, P. H.; Holms, A. G.; Davison, H. W.

1973-01-01

The maximum fuel pin temperature for nominal operation in an advanced power reactor is 1370 K. Because of possible nitrogen embrittlement of the clad, the fuel temperature was limited to 1622 K. Assuming simultaneous occurrence of the most adverse conditions a deterministic analysis gave a maximum fuel temperature of 1610 K. A statistical analysis, using a synthesized estimate of the standard deviation for the highest fuel pin temperature, showed probabilities of 0.015 of that pin exceeding the temperature limit by the distribution free Chebyshev inequality and virtually nil assuming a normal distribution. The latter assumption gives a 1463 K maximum temperature at 3 standard deviations, the usually assumed cutoff. Further, the distribution and standard deviation of the fuel-clad gap are the most significant contributions to the uncertainty in the fuel temperature.

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

Development of 3D pseudo pin-by-pin calculation methodology in ANC

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

Zhang, B.; Mayhue, L.; Huria, H.

2012-07-01

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000{sup R} plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. Themore » mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)« less

Initial Microstructure Evaluation of a U3Si2 + W Fuel Pin Fabricated Via Arc Melt Gravity Drop Casting

NASA Astrophysics Data System (ADS)

Hoggan, Rita E.; Harp, Jason M.

2018-02-01

Injection casting has historically been used to fabricate metallic nuclear fuel on a large scale. Casting of intermetallic fuel forms, such as U3Si2, may be an alternative pathway for fabrication of fuel pins to powder metallurgy. To investigate casting on a small scale, arc melt gravity drop casting was employed to cast a one-off pin of U3Si2 for evaluation as a fabrication method for U3Si2 as a light water reactor fuel. The pin was sectioned and examined via optical microscopy and scanning electron microscopy equipped with energy dispersive x-ray spectroscopy (EDS). Image analysis was used to estimate the volume fraction of phase impurities as well as porosity. The primary phase determined by EDS was U3Si2 with U-O and U-Si-W phase impurities. Unusually high levels of tungsten were observed because of accidental tungsten introduction during arc melting. No significant changes in microstructure were observed after annealing a section of the pin at 800°C for 72 h. The average density of the sectioned specimens was 12.4 g/cm3 measured via Archimedes principle immersion density and He gas displacement.

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

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

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

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

Posttest examination of Sodium Loop Safety Facility experiments. [LMFBR

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

Holland, J.W.

In-reactor, safety experiments performed in the Sodium Loop Safety Facility (SLSF) rely on comprehensive posttest examinations (PTE) to characterize the postirradiation condition of the cladding, fuel, and other test-subassembly components. PTE information and on-line instrumentation data, are analyzed to identify the sequence of events and the severity of the accident for each experiment. Following in-reactor experimentation, the SLSF loop and test assembly are transported to the Hot Fuel Examination Facility (HFEF) for initial disassembly. Goals of the HFEF-phase of the PTE are to retrieve the fuel bundle by dismantling the loop and withdrawing the test assembly, to assess the macro-conditionmore » of the fuel bundle by nondestructive examination techniques, and to prepare the fuel bundle for shipment to the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory.« less

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

Enquiry into the Topology of Plasma Membrane-Localized PIN Auxin Transport Components.

PubMed

Nodzyński, Tomasz; Vanneste, Steffen; Zwiewka, Marta; Pernisová, Markéta; Hejátko, Jan; Friml, Jiří

2016-11-07

Auxin directs plant ontogenesis via differential accumulation within tissues depending largely on the activity of PIN proteins that mediate auxin efflux from cells and its directional cell-to-cell transport. Regardless of the developmental importance of PINs, the structure of these transporters is poorly characterized. Here, we present experimental data concerning protein topology of plasma membrane-localized PINs. Utilizing approaches based on pH-dependent quenching of fluorescent reporters combined with immunolocalization techniques, we mapped the membrane topology of PINs and further cross-validated our results using available topology modeling software. We delineated the topology of PIN1 with two transmembrane (TM) bundles of five α-helices linked by a large intracellular loop and a C-terminus positioned outside the cytoplasm. Using constraints derived from our experimental data, we also provide an updated position of helical regions generating a verisimilitude model of PIN1. Since the canonical long PINs show a high degree of conservation in TM domains and auxin transport capacity has been demonstrated for Arabidopsis representatives of this group, this empirically enhanced topological model of PIN1 will be an important starting point for further studies on PIN structure-function relationships. In addition, we have established protocols that can be used to probe the topology of other plasma membrane proteins in plants. Copyright © 2016 The Authors. Published by Elsevier Inc. All rights reserved.

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

Popov, Emilian L.; Pointer, William David

This work assesses the influence of assumptions made when generating a mesh of a wire-wrappedgeometry. The contact region between a wire and its adjacent pin is commonly modeled by eitherembedding the wire to the adjacent pin or trimming the wire so that a gap separates the wire from itsadjacent pin. These models are referred to as close-gap and open-gap approaches herein and are applied totwo geometries. The first geometry consists of a single pin wire-wrapped subchannel. A polyhedral meshand a hexahedral mesh are generated. The second and third geometry are a 7-pin and a 19-pinwire-wrapped bundles meshed with polyhedral elementsmore » only. Pressure drops are obtained with theSTAR-CCM+computational fluid dynamic package. Sensitivity analyses of the mesh density, the meshtype, and the turbulent models are performed. Numerical results show that the best match to theexperimental data and to the Cheng-Todreas correlation is obtained with the combination of a hexahedralmesh, the shear stress transport (SST) turbulent model, and the open-gap approach. In the case of the 7-pingeometry, the best results are obtained with the open-gap approach and the SST turbulent model. The19-pin geometry yields contradictory results to the 7-pin geometry results, and thus will require furtherinvestigations.« less

Expression of proteinase inhibitor II proteins during floral development in Solanum americanum.

PubMed

Sin, Suk-Fong; Chye, Mee-Len

2004-10-01

The heterologous expression of serine proteinase inhibitor II (PIN2) proteins confers insect resistance in transgenic plants, but little is known of their endogenous roles. We have cloned two cDNAs encoding Solanum americanum PIN2 proteins, SaPIN2a and SaPIN2b. SaPIN2a is highly expressed in stem, particularly in the phloem, suggesting it could possibly regulate proteolysis in the sieve elements. When SaPIN2a was expressed in transgenic lettuce, we observed an inhibition of endogenous trypsin- and chymotrypsin-like activities. Here, we demonstrate that both SaPIN2a and SaPIN2b are expressed in floral tissues that are destined to undergo developmental programmed cell death (PCD), suggesting possible endogenous roles in inhibiting trypsin- and chymotrypsin-like activities during flower development. Northern and western blot analyses revealed that SaPIN2a and SaPIN2b mRNAs and proteins show highest expression early in floral development. In situ hybridization analysis and immunolocalization on floral sections, localized SaPIN2a and SaPIN2b mRNAs and their proteins to tissues that would apparently undergo PCD: the ovules, the stylar transmitting tissue, the stigma and the vascular bundles. Detection of PCD in floral sections was achieved using terminal deoxynucleotidyltransferase-mediated dUTP nick end labeling (TUNEL) analysis. Examination of the mid-style before, and 1 day after, pollination revealed that high expression of SaPIN2a and SaPIN2b in the style was inversely correlated with PCD.

Material distribution in light water reactor-type bundles tested under severe accident conditions

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

Noack, V.; Hagen, S.J.L.; Hofmann, P.

1997-02-01

Severe fuel damage experiments simulating small-break loss-of-coolant accidents have been carried out in the CORA out-of-pile test facility at Forschungszentrum Karlsruhe. Rod bundles with electrically heated fuel rod simulators containing annular UO{sub 2} pellets, UO{sub 2} full pellet rods, and absorber rods of two kinds (Ag/In/Cd to represent pressurized water reactor conditions and B{sub 4}C to represent boiling water reactor and VVER-1000 fuel elements) were subjected to temperature transients up to 2,300 K. A special method was applied to determine the axial mass distribution of bundle materials. The low-temperature melt formation by various interactions between zirconium and components of absorbermore » and spacer grids strongly influences the bundle degradation and material relocation. Absorber materials can separate from the fuel by a noncoherent relocation of the materials at different temperatures. The distributions of solidified materials in the different test bundles show a clear dependence on the axial temperature profile. Coolant channel blockages are observed mainly at the lower end of the bundle, i.e., near the lowest elevation at which an oxidation excursion resulting from the highly exothermic zirconium-steam reaction had been experienced. This elevation corresponds with a steep axial temperature gradient in the maximum temperature attained. Oxide layers on Zircaloy result in reduced melt formation.« less

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

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.

Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

DOEpatents

Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

2012-04-24

A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

Use of freeze-casting in advanced burner reactor fuel design

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

Lang, A. L.; Yablinsky, C. A.; Allen, T. R.

2012-07-01

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by thatmore » fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results show that criticality is achievable with freeze-cast fuel pins despite the significant amount of inert fuel matrix. Freeze casting is a promising method to achieve very precise fuel placement within fuel pins. (authors)« less

Safety consequences of local initiating events in an LMFBR

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

Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.

1975-12-01

The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered.

Tribological Properties Of Coal Slurries

NASA Technical Reports Server (NTRS)

Fusaro, Robert L.; Schrubens, Dale L.

1988-01-01

Report describes study of tribological properties of coal/methanol slurries with pin-on-disk tribometer. Coefficients of friction, rates of wear of steel pin, and morphological studies of worn surfaces conducted on pins and disks of AISI 440C HT stainless steel and M-50 tool steel, both used as bearing steels. Coal slurries considered as replacement fuels in terrestrial oil-burning facilities and possible fuels for future aircraft turbine engines. Rates of wear of metallic components through which slurries flow limit such practical applications.

Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

NASA Astrophysics Data System (ADS)

Grande, Lisa Christine

A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

A microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

SCAN+

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

Kenneth Krebs, John Svoboda

2009-11-01

SCAN+ is a software application specifically designed to control the positioning of a gamma spectrometer by a two dimensional translation system above spent fuel bundles located in a sealed spent fuel cask. The gamma spectrometer collects gamma spectrum information for the purpose of spent fuel cask fuel loading verification. SCAN+ performs manual and automatic gamma spectrometer positioning functions as-well-as exercising control of the gamma spectrometer data acquisitioning functions. Cask configuration files are used to determine the positions of spent fuel bundles. Cask scanning files are used to determine the desired scan paths for scanning a spent fuel cask allowing formore » automatic unattended cask scanning that may take several hours.« less

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.

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

A fission gas release correlation for uranium nitride fuel pins

NASA Technical Reports Server (NTRS)

Weinstein, M. B.; Davison, H. W.

1973-01-01

A model was developed to predict fission gas releases from UN fuel pins clad with various materials. The model was correlated with total release data obtained by different experimentors, over a range of fuel temperatures primarily between 1250 and 1660 K, and fuel burnups up to 4.6 percent. In the model, fission gas is transported by diffusion mechanisms to the grain boundaries where the volume grows and eventually interconnects with the outside surface of the fuel. The within grain diffusion coefficients are found from fission gas release rate data obtained using a sweep gas facility.

Measurement of carbon distribution in nuclear fuel pin cladding specimens by means of a secondary ion mass spectrometer

NASA Astrophysics Data System (ADS)

Bart, Gerhard; Aerne, Ernst Tino; Burri, Martin; Zwicky, Hans-Urs

1986-11-01

Cladding carburization during irradiation of advanced mixed uranium plutonium carbide fast breeder reactor fuel is possibly a life limiting fuel pin factor. The quantitative assessment of such clad carbon embrittlement is difficult to perform by electron microprobe analysis because of sample surface contamination, and due to the very low energy of the carbon K α X-ray transition. The work presented here describes a method developed at the Swiss Federal Institute for Reactor Research (EIR) to use shielded secondary ion mass spectrometry (SIMS) as an accurate tool to determine radial distribution profiles of carbon in radioactive stainless steel fuel pin cladding. Compared with nuclear microprobe analysis (NMA) [1], which is also an accurate method for carbon analysis, the SIMS method distinguishes itself by its versatility for simultaneous determination of additional impurities.

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

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

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

2011-09-01

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

Development of burnup dependent fuel rod model in COBRA-TF

NASA Astrophysics Data System (ADS)

Yilmaz, Mine Ozdemir

The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN predictions. After confirming that the new fuel thermal conductivity model in CTF worked and provided consistent results with FRAPTRAN predictions for a single fuel rod configuration, the same type of analysis was carried out for a bigger system which is the 4x4 PWR bundle consisting of 15 fuel pins and one control guide tube. Steady- state calculations at Hot Full Power (HFP) conditions for control guide tube out (unrodded) were performed using the 4x4 PWR array with CTF/TORT-TD coupled code system. Fuel centerline, surface and average temperatures predicted by CTF/TORT-TD with and without the new fuel thermal conductivity model were compared against CTF/TORT-TD/FRAPTRAN predictions to demonstrate the improvement in fuel centerline predictions when new model was used. In addition to that constant and CTF dynamic gap conductance model were used with the new thermal conductivity model to show the performance of the CTF dynamic gap conductance model and its impact on fuel centerline and surface temperatures. Finally, a Rod Ejection Accident (REA) scenario using the same 4x4 PWR array was run both at Hot Zero Power (HZP) and Hot Full Power (HFP) condition, starting at a position where half of the control rod is inserted. This scenario was run using CTF/TORT-TD coupled code system with and without the new fuel thermal conductivity model. The purpose of this transient analysis was to show the impact of thermal conductivity degradation (TCD) on feedback effects, specifically Doppler Reactivity Coefficient (DRC) and, eventually, total core reactivity.

Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

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

Daniel, W. E.; Rudisill, T. S.; O'Rourke, P. E.

2017-01-26

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgasmore » composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.« less

Performance of low smeared density sodium-cooled fast reactor metal fuel

NASA Astrophysics Data System (ADS)

Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

2015-10-01

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

CHF considerations for highly moderated 100% MOX fuels PWRs

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

Saphier, D.; Raymond, P.

1995-09-01

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

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.

Unrestrained swelling of uranium-nitride fuel irradiated at temperatures ranging from 1100 to 1400 K (1980 to 2520 R)

NASA Technical Reports Server (NTRS)

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

1973-01-01

Six fuel pins were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a stainless steel (type 304L) clad. The pins were irradiated for approximately 4000 hours to burnups of about 2.0 atom percent uranium. The average clad surface temperature during irradiation was about 1100 K (1980 deg R). Since stainless steel has a very low creep strength relative to that of UN at this temperature, these tests simulated unrestrained swelling of UN. The tests indicated that at 1 percent uranium atom burnup the unrestrained diametrical swelling of UN is about 0.5, 0.8, and 1.0 percent at 1223, 1264, and 1306 K (2200, deg 2273 deg, and 2350 deg R), respectively. The tests also indicated that the irradiation induced swelling of unrestrained UN fuel pellets appears to be isotropic.

76 FR 62653 - Airworthiness Directives; Airbus Model A310 Series Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-11

... the fuel electrical circuit in the Right Hand (RH) wing must be modified in order to ensure better... 2S of the fuel electrical circuit in the Right Hand (RH) wing must be modified in order to ensure... the wire bundle 2S in the RH wing pylon area to the generator wire bundle of engine 2. The...

Evaluation of COBRA III-C and SABRE-I (wire wrap version) computational results by comparison with steady-state data from a 19-pin internally guard heated sodium cooled bundle with a six-channel central blockage (THORS bundle 3C). [LMFBR

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

Dearing, J F; Rose, S D; Nelson, W R

The predicted computational results of two well-known sub-channel analysis codes, COBRA-III-C and SABRE-I (wire wrap version), have been evaluated by comparison with steady state temperature data from the THORS Facility at ORNL. Both codes give good predictions of transverse and axial temperatures when compared with wire wrap thermocouple data. The crossflow velocity profiles predicted by these codes are similar which is encouraging since the wire wrap models are based on different assumptions.

Performance of low smeared density sodium-cooled fast reactor metal fuel

DOE PAGES

Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; ...

2015-06-17

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactormore » designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.« less

Development of High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Farmer, J.; Dixon, D.; Kapernick, R.; Dickens, R.; Adams, M.

2007-01-01

Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Work at the NASA Marshall Space Flight Center seeks to develop high fidelity thermal simulators that not only match the static power profile that would be observed in an operating, fueled nuclear reactor, but to also match the dynamic fuel pin performance during feasible transients. Comparison between the fuel pins and thermal simulators is made at the fuel clad surface, which corresponds to the sheath surface in the thermal simulator. Static and dynamic fuel pin performance was determined using SINDA-FLUINT analysis, and the performance of conceptual thermal simulator designs was compared to the expected nuclear performance. Through a series of iterative analysis, a conceptual high fidelity design will be developed, followed by engineering design, fabrication, and testing to validate the overall design process. Although the resulting thermal simulator will be designed for a specific reactor concept, establishing this rigorous design process will assist in streamlining the thermal simulator development for other reactor concepts.

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

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

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

2002-05-15

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

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

Target-fueled nuclear reactor for medical isotope production

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

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

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

Luna, R. E.

This paper provides a simple model for estimating the release of respirable aerosols resulting from an attack on a spent fuel cask using a high energy density device (HEDD). Two primary experiments have provided data on potential releases from spent fuel casks under HEDD attack. Sandia National Laboratories (SNL) conducted the first in the early 1980's and the second was sponsored by Gessellshaft fur Anlagen- and Reaktorsicherheit (GRS) in Germany and conducted in France in 1994. Both used surrogate spent fuel assemblies in real casks. The SNL experiments used un-pressurized fuel pin assemblies in a single element cask while themore » GRS tests used pressurized fuel pin assemblies in a 9-element cask. Data from the two test programs is reasonably consistent, given the differences in the experiments, but the use of the test data for prediction of releases resulting from HEDD attack requires a method for accounting for the effects of pin pressurization release and the ratio of pin plenum gas release to cask free volume (VR). To account for the effects of VR and to link the two data sources, a simple model has been developed that uses both the SNL data and the GRS data as well as recent test data on aerosols produced in experiments with single pellets subjected to HEDD effects conducted under the aegis of the International Consortium's Working Group on Sabotage of Transport and Storage Casks (WGSTSC). (authors)« less

PWR integral tie plate and locking mechanism

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

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

1980-08-26

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

Quasi-heterogeneous efficient 3-D discrete ordinates CANDU calculations using Attila

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

Preeti, T.; Rulko, R.

2012-07-01

In this paper, 3-D quasi-heterogeneous large scale parallel Attila calculations of a generic CANDU test problem consisting of 42 complete fuel channels and a perpendicular to fuel reactivity device are presented. The solution method is that of discrete ordinates SN and the computational model is quasi-heterogeneous, i.e. fuel bundle is partially homogenized into five homogeneous rings consistently with the DRAGON code model used by the industry for the incremental cross-section generation. In calculations, the HELIOS-generated 45 macroscopic cross-sections library was used. This approach to CANDU calculations has the following advantages: 1) it allows detailed bundle (and eventually channel) power calculationsmore » for each fuel ring in a bundle, 2) it allows the exact reactivity device representation for its precise reactivity worth calculation, and 3) it eliminates the need for incremental cross-sections. Our results are compared to the reference Monte Carlo MCNP solution. In addition, the Attila SN method performance in CANDU calculations characterized by significant up scattering is discussed. (authors)« less

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Kazimi, Mujid S.

2013-10-01

The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

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

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

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

1981-04-01

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

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

DOE PAGES

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

2016-02-27

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

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

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

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

2016-05-01

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

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

NASA Astrophysics Data System (ADS)

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

2016-05-01

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

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

Wilson, C.L.; Hesson, G.M.; Pilger, J.P.

1993-09-01

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuelmore » bundle is cooled.« less

Tin-oxide-coated single-walled carbon nanotube bundles supporting platinum electrocatalysts for direct ethanol fuel cells.

PubMed

Hsu, Ryan S; Higgins, Drew; Chen, Zhongwei

2010-04-23

Novel tin-oxide (SnO(2))-coated single-walled carbon nanotube (SWNT) bundles supporting platinum (Pt) electrocatalysts for ethanol oxidation were developed for direct ethanol fuel cells. SnO(2)-coated SWNT (SnO(2)-SWNT) bundles were synthesized by a simple chemical-solution route. SnO(2)-SWNT bundles supporting Pt (Pt/SnO(2)-SWNTs) electrocatalysts and SWNT-supported Pt (Pt/SWNT) electrocatalysts were prepared by an ethylene glycol reduction method. The catalysts were physically characterized using TGA, XRD and TEM and electrochemically evaluated through cyclic voltammetry experiments. The Pt/SnO(2)-SWNTs showed greatly enhanced electrocatalytic activity for ethanol oxidation in acid medium, compared to the Pt/SWNT. The optimal SnO(2) loading of Pt/SnO(2)-SWNT catalysts with respect to specific catalytic activity for ethanol oxidation was also investigated.

Critical current and flux dynamics in Ag-doped FeSe superconductor

NASA Astrophysics Data System (ADS)

Galluzzi, A.; Polichetti, M.; Buchkov, K.; Nazarova, E.; Mancusi, D.; Pace, S.

2017-02-01

The measurements of DC magnetization as a function of the temperature M(T), magnetic field M(H), and time M(t) have been performed in order to compare the superconducting and pinning properties of an undoped FeSe0.94 sample and a silver doped FeSe0.94 + 6 wt% Ag sample. The M(T) curves indicate an improvement of the superconducting critical temperature and a reduction of the non-superconducting phase Fe7Se8 due to the silver doping. This is confirmed by the field and temperature dependent critical current density Jc(H,T) extracted from the superconducting hysteresis loops at different temperatures within the Bean critical state model. Moreover, the combined analysis of the Jc(T) and of the pinning force Fp(H/Hirr) indicate that the pinning mechanisms in both samples can be described in the framework of the collective pinning theory. The U*(T, J) curves show a pinning crossover from an elastic creep regime of intermediate size flux bundles, for low temperatures, to a plastic creep regime at higher temperatures for both the samples. Finally, the vortex hopping attempt time has been evaluated for both samples and the results are comparable with the values reported in the literature for high Tc materials.

High energy X-ray CT study on the central void formations and the fuel pin deformations of FBR fuel assemblies

NASA Astrophysics Data System (ADS)

Katsuyama, Kozo; Nagamine, Tsuyoshi; Matsumoto, Shin-ichiro; Sato, Seichi

2007-02-01

The central void formations and deformations of fuel pins were investigated in fuel assemblies irradiated to high burn-up, using a non-destructive X-ray CT (computer tomography) technique. In this X-ray CT, the effect of strong gamma ray activity could be reduced to a negligible degree by using the pulse of a high energy X-ray source and detecting the intensity of the transmitted X-rays in synchronization with the generated X-rays. Clear cross-sectional images of fuel assemblies irradiated to high burn-up in a fast breeder reactor were successively obtained, in which the wrapping wires, cladding, pellets and central voids could be distinctly seen. The diameter of a typical central void measured by X-ray CT agreed with the one obtained by ceramography within an error of 0.1 mm. Based on this result, the dependence of the central void diameter on the linear heating rate was analyzed. In addition, the deformation behavior of a fuel pin along its axial direction could be analyzed from 20 stepwise X-ray cross-sectional images obtained in a small interval, and the results obtained showed a good agreement with the predictions calculated by two computer codes.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

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

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-02-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish releasemore » fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.« less

PRELIMINARY COUPLING OF THE MONTE CARLO CODE OPENMC AND THE MULTIPHYSICS OBJECT-ORIENTED SIMULATION ENVIRONMENT (MOOSE) FOR ANALYZING DOPPLER FEEDBACK IN MONTE CARLO SIMULATIONS

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

Matthew Ellis; Derek Gaston; Benoit Forget

In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes.more » An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.« less

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

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

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

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

Interlocking egg-crate type grid assembly

DOEpatents

Kast, Steven J.

1987-01-01

Disclosed is an interlocking egg-crate hexagonal grid for supporting a nuclear fuel pin in a hexagonal array. The grid is formed from strips bent at an angle of about 120.degree. at each vertex. Over some faces of each hexagonal cell the strips are coplanar but are arranged, by stacking and interlocking, to avoid any double thickness of metal in that plane. Springs and dimples are formed in the faces of each cell to hold the fuel pin substantially centered.

Interlocking egg-crate type grid assembly

DOEpatents

Kast, S.J.

1985-03-15

Disclosed is an interlocking egg-crate hexagonal grid for supporting a nuclear fuel pin in a hexagonal array. The grid is formed from strips bent at an angle of about 120/sup 0/ at each vertex. Over some faces of each hexagonal cell the strips are coplanar but are arranged, by stacking interlocking, to avoid any double thickness of metal in that plane. Springs and dimples are formed in the faces of each cell to hold the fuel pin substantially centered.

Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

DOE PAGES

Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher L.; ...

2016-04-29

In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may findmore » other safety and security or safeguards applications, such as arms control verification.« less

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

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

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara

2005-02-06

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an earlymore » prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.« less

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.

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

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

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

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

S. Bays; W. Skerjanc; M. Pope

A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch averagemore » discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.« less

Fuel pins with both target and fuel pellets in an isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

Fuel pin cladding

DOEpatents

Vaidyanathan, Swaminathan; Adamson, Martyn G.

1986-01-01

An improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

DOE PAGES

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena; ...

2016-10-22

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

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

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

NASA Astrophysics Data System (ADS)

Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A. A.

2017-01-01

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

Welding fixture for nuclear fuel pin cladding assemblies

DOEpatents

Oakley, David J.; Feld, Sam H.

1986-01-01

A welding fixture for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

Decontamination apparatus and method. [Patent applications

DOEpatents

Oakley, D.J.

1983-12-16

This invention relates generally to the fabrication of fuel pin elements employed in nuclear reactors and, more particularly, to removing radioactive contamination disposed on the exterior of finally assembled fuel pins. A blast head includes a plurality of spray nozzles mounted in a chamber for receiving a workpiece. The several spray nozzles concurrently direct a plurality of streams of a pressurized gas and abrasive grit mixture toward a peripheral portion of the workpiece to remove particulates or debris therefrom. An exhaust outlet is formed in the chamber for discharging the particulates and spent grit.

77 FR 2658 - Airworthiness Directives; Bombardier, Inc. Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-19

... high pressure fuel lines due to improper installation of an expandable pin on the lower cowl assembly... chafing of the high pressure fuel lines, which if not corrected, could cause fuel leakage in a fire zone... on a high pressure (HP) fuel line. The source of chafing was related to the improper installation of...

Numerical Simulations on the Laser Spot Welding of Zirconium Alloy Endplate for Nuclear Fuel Bundle Assembly

NASA Astrophysics Data System (ADS)

Satyanarayana, G.; Narayana, K. L.; Boggarapu, Nageswara Rao

2018-03-01

In the nuclear industry, a critical welding process is joining of an end plate to a fuel rod to form a fuel bundle. Literature on zirconium welding in such a critical operation is limited. A CFD model is developed and performed for the three-dimensional non-linear thermo-fluid analysis incorporating buoyancy and Marnangoni stress and specifying temperature dependent properties to predict weld geometry and temperature field in and around the melt pool of laser spot during welding of a zirconium alloy E110 endplate with a fuel rod. Using this method, it is possible to estimate the weld pool dimensions for the specified laser power and laser-on-time. The temperature profiles will estimate the HAZ and microstructure. The adequacy of generic nature of the model is validated with existing experimental data.

Sensitivity Analysis of OECD Benchmark Tests in BISON

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

Swiler, Laura Painton; Gamble, Kyle; Schmidt, Rodney C.

2015-09-01

This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on sensitivity analysis of a fuels performance benchmark problem. The benchmark problem was defined by the Uncertainty Analysis in Modeling working group of the Nuclear Science Committee, part of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD ). The benchmark problem involv ed steady - state behavior of a fuel pin in a Pressurized Water Reactor (PWR). The problem was created in the BISON Fuels Performance code. Dakota was used to generate and analyze 300 samples of 17 input parameters defining coremore » boundary conditions, manuf acturing tolerances , and fuel properties. There were 24 responses of interest, including fuel centerline temperatures at a variety of locations and burnup levels, fission gas released, axial elongation of the fuel pin, etc. Pearson and Spearman correlatio n coefficients and Sobol' variance - based indices were used to perform the sensitivity analysis. This report summarizes the process and presents results from this study.« less

Refractory oxide insulated thermocouple designed and analyzed for high temperature applications

NASA Technical Reports Server (NTRS)

Popper, G. F.; Zeren, T. Z.

1969-01-01

Study establishes design criteria for constructing high temperature thermocouple to measure nuclear fuel pin temperature. The study included a literature search to determine the compatibility of material useful for thermocouples, a hot zone error analysis, and a prototype design for hot junction and connector pin connections.

Creating NDA working standards through high-fidelity spent fuel modeling

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

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

2012-01-01

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

System and method for reducing combustion dynamics in a combustor

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward; Zuo, Baifang; York, William David

2015-09-01

A system for reducing combustion dynamics in a combustor includes an end cap having an upstream surface axially separated from a downstream surface, and tube bundles extend from the upstream surface through the downstream surface. A divider inside a tube bundle defines a diluent passage that extends axially through the downstream surface, and a diluent supply in fluid communication with the divider provides diluent flow to the diluent passage. A method for reducing combustion dynamics in a combustor includes flowing a fuel through tube bundles, flowing a diluent through a diluent passage inside a tube bundle, wherein the diluent passage extends axially through at least a portion of the end cap into a combustion chamber, and forming a diluent barrier in the combustion chamber between the tube bundle and at least one other adjacent tube bundle.

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

NASA Astrophysics Data System (ADS)

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

2017-04-01

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

Evaluation of Future Fuels in a High Pressure Common Rail System. Part 3. John Deere 4.5L Powertech Plus

DTIC Science & Technology

2013-01-01

within the fuel injectors regardless of viscosity or lubricity levels. 15. SUBJECT TERMS Fuel Lubricity, Viscosity, HPCR, Synthetic Jet Fuel, Denso... injectors regardless of viscosity or lubricity levels. UNCLASSIFIED v UNCLASSIFIED FOREWORD/ACKNOWLEDGMENTS The U.S. Army TARDEC...40 3.10 UPPER INJECTOR CONNECTING PIN

Fuel pin cladding

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1986-01-28

Disclosed is an improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients. 2 figs.

Fuel pin cladding

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1983-12-16

An improved fuel pin cladding, particularly adapted for use in breeder reactors, is described which consist of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel an/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

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

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

Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times.

An engineering closure for heavily under-resolved coarse-grid CFD in large applications

NASA Astrophysics Data System (ADS)

Class, Andreas G.; Yu, Fujiang; Jordan, Thomas

2016-11-01

Even though high performance computation allows very detailed description of a wide range of scales in scientific computations, engineering simulations used for design studies commonly merely resolve the large scales thus speeding up simulation time. The coarse-grid CFD (CGCFD) methodology is developed for flows with repeated flow patterns as often observed in heat exchangers or porous structures. It is proposed to use inviscid Euler equations on a very coarse numerical mesh. This coarse mesh needs not to conform to the geometry in all details. To reinstall physics on all smaller scales cheap subgrid models are employed. Subgrid models are systematically constructed by analyzing well-resolved generic representative simulations. By varying the flow conditions in these simulations correlations are obtained. These comprehend for each individual coarse mesh cell a volume force vector and volume porosity. Moreover, for all vertices, surface porosities are derived. CGCFD is related to the immersed boundary method as both exploit volume forces and non-body conformal meshes. Yet, CGCFD differs with respect to the coarser mesh and the use of Euler equations. We will describe the methodology based on a simple test case and the application of the method to a 127 pin wire-wrap fuel bundle.

A Modeling Approach to Fiber Fracture in Melt Impregnation

NASA Astrophysics Data System (ADS)

Ren, Feng; Zhang, Cong; Yu, Yang; Xin, Chunling; Tang, Ke; He, Yadong

2017-02-01

The effect of process variables such as roving pulling speed, melt temperature and number of pins on the fiber fracture during the processing of thermoplastic based composites was investigated in this study. The melt impregnation was used in this process of continuous glass fiber reinforced thermoplastic composites. Previous investigators have suggested a variety of models for melt impregnation, while comparatively little effort has been spent on modeling the fiber fracture caused by the viscous resin. Herein, a mathematical model was developed for impregnation process to predict the fiber fracture rate and describe the experimental results with the Weibull intensity distribution function. The optimal parameters of this process were obtained by orthogonal experiment. The results suggest that the fiber fracture is caused by viscous shear stress on fiber bundle in melt impregnation mold when pulling the fiber bundle.

Welding fixture for nuclear fuel pin cladding assemblies

DOEpatents

Oakley, D.J.; Feld, S.H.

1984-02-22

A welding fixture is described for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

Electrolytic/fuel cell bundles and systems including a current collector in communication with an electrode thereof

DOEpatents

Hawkes, Grant L.; Herring, James S.; Stoots, Carl M.; O& #x27; Brien, James E.

2013-03-05

Electrolytic/fuel cell bundles and systems including such bundles include an electrically conductive current collector in communication with an anode or a cathode of each of a plurality of cells. A cross-sectional area of the current collector may vary in a direction generally parallel to a general direction of current flow through the current collector. The current collector may include a porous monolithic structure. At least one cell of the plurality of cells may include a current collector that surrounds an outer electrode of the cell and has at least six substantially planar exterior surfaces. The planar surfaces may extend along a length of the cell, and may abut against a substantially planar surface of a current collector of an adjacent cell. Methods for generating electricity and for performing electrolysis include flowing current through a conductive current collector having a varying cross-sectional area.

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

Yuuichi Tooya; Tadahiro Washiya; Kenji Koizumi

Japan Atomic Energy Agency (JAEA) has been leading feasibility study on commercialized fast reactor cycle systems in Japan. In this study, we have proposed a new disassembly technology by mechanical disassembly system that consists of a mechanical cutting step and a wrapper tube pulling step. In the mechanical disassembly system, high durability mechanical tool grinds the wrapper tube (Slit-cut (S/C) operation in circle direction), and then the wrapper tube is pulled out and removed from the fuel assembly. Then the fuel pins are cut (Crop-cut (C/C) operation at entrance nozzle side) and the entrance nozzle is removed. The fuel pinsmore » are transported to the shearing device in next process. The Fundamental tests were carried out with simulated FBR fuel pins and wrapper tube, and cutting performance and wrapper tube pulling performance has been confirmed by engineering scale. As results, we established an efficient disassembly procedure and the fundamental design of mechanical disassembly system. (authors)« less

Detecting pin diversion from pressurized water reactors spent fuel assemblies

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

Ham, Young S.; Sitaraman, Shivakumar

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

78 FR 16198 - Airworthiness Directives; The Boeing Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-14

... determine if wires touch the upper surface of the center upper auxiliary fuel tank, and marking the location, as necessary; inspecting all wire bundles above the center upper auxiliary fuel tank for splices and... requires inspecting to determine if wires touch the upper surface of the center upper auxiliary fuel tank...

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors. FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.

Benchmarking of calculation schemes in APOLLO2 and COBAYA3 for WER lattices

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

Zheleva, N.; Ivanov, P.; Todorova, G.

This paper presents solutions of the NURISP WER lattice benchmark using APOLLO2, TRIPOLI4 and COBAYA3 pin-by-pin. The main objective is to validate MOC based calculation schemes for pin-by-pin cross-section generation with APOLLO2 against TRIPOLI4 reference results. A specific objective is to test the APOLLO2 generated cross-sections and interface discontinuity factors in COBAYA3 pin-by-pin calculations with unstructured mesh. The VVER-1000 core consists of large hexagonal assemblies with 2 mm inter-assembly water gaps which require the use of unstructured meshes in the pin-by-pin core simulators. The considered 2D benchmark problems include 19-pin clusters, fuel assemblies and 7-assembly clusters. APOLLO2 calculation schemes withmore » the step characteristic method (MOC) and the higher-order Linear Surface MOC have been tested. The comparison of APOLLO2 vs. TRIPOLI4 results shows a very close agreement. The 3D lattice solver in COBAYA3 uses transport corrected multi-group diffusion approximation with interface discontinuity factors of Generalized Equivalence Theory (GET) or Black Box Homogenization (BBH) type. The COBAYA3 pin-by-pin results in 2, 4 and 8 energy groups are close to the reference solutions when using side-dependent interface discontinuity factors. (authors)« less

Experimental study of burnout in channels with twisted fuel rods

NASA Astrophysics Data System (ADS)

Bol'Shakov, V. V.; Bashkirtsev, S. M.; Kobzar', L. L.; Morozov, A. G.

2007-05-01

The results of experimental studies of pressure drop and critical heat flux in the models of fuel assemblies (FAs) with fuel rod simulators twisted relative to the longitudinal axis and a three-ray cross section are considered. The experimental data are compared to the results obtained with the use of techniques adopted for design calculations with fuel rod bundles of type-VVER reactors.

Coupling procedure for TRANSURANUS and KTF codes

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

Jimenez, J.; Alglave, S.; Avramova, M.

2012-07-01

The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned toolsmore » can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)« less

Development of self-interrogation neutron resonance densitometry (sinrd) to measure the fissile content in nuclear fuel

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne Marie

The development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: (1) SINRD provides absolute measurements of burnup independent of the operator's declaration. (2) SINRD is sensitive to pin removal over the entire burnup range and can verify the diversion of 6% of fuel pins within 3o from LWR spent LEU and MOX fuel. (3) SINRD is insensitive to the boron concentration and initial fuel enrichment and can therefore be used at multiple spent fuel storage facilities. (4) The calibration of SINRD at one reactor facility carries over to reactor sites in different countries because it uses the ratio of fission chambers (FCs) that are not facility dependent. (5) SINRD can distinguish fresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent LEU fuel without using reactor burnup codes.

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

Barber, Laura Jeaneen

The purpose of the L’Innovator is to assemble unique, state-of-the-art IP bundles developed at the national labs that aggregate synergistic technologies in furtherance of the emerging hydrogen and fuel cell market. The first L’Innovator IP bundle consists of Oxygen Reduction Reaction (ORR) Catalyst technology developed at Brookhaven National Laboratory (BNL), combined with Membrane Electrode Assembly (MEA) technology developed at Los Alamos National Laboratory (LANL).

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

NASA Astrophysics Data System (ADS)

Mella, R.; Wenman, M. R.

2013-06-01

Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of different FE and finite difference models. Non-linear mechanical behaviour of the fuel and cladding including, fuel creep and swelling and cladding creep and plasticity each with dependencies on a variety of different properties. A fission gas release model which takes inputs from first principles calculations. Explicitly integrated inventory calculations performed in a coupled manner. Freedom to model steady state and transient behaviour using implicit time integration. The whole pin geometry is considered over an entire typical fuel life. The model showed by examination of normal operation and a subsequent transient chosen for software demonstration purposes: ABAQUS may be a sufficiently flexible platform to develop a complete and verified fuel performance code. The importance and effectiveness of the geometry of the fuel spacer pellets was characterised. The fuels performance under normal conditions (high friction no power spikes) would not suggest serious degradation of the cladding in fuel life. Large plastic strains were found when pellet bonding was strong, these would appear at all pellets cladding triple points and all pellet radial crack and cladding interfaces thus showing a possible axial direction to cracks forming from ductility exhaustion.

Molybdenum-UO2 cermet irradiation at 1145 K.

NASA Technical Reports Server (NTRS)

Mcdonald, G.

1971-01-01

Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

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.

Prise en compte d'un couplage fin neutronique-thermique dans les calculs d'assemblage pour les reacteurs a eau pressurisee

NASA Astrophysics Data System (ADS)

Greiner, Nathan

Core simulations for Pressurized Water Reactors (PWR) is insured by a set of computer codes which allows, under certain assumptions, to approximate the physical quantities of interest, such as the effective multiplication factor or the power or temperature distributions. The neutronics calculation scheme relies on three great steps : -- the production of an isotopic cross-sections library ; -- the production of a reactor database through the lattice calculation ; -- the full-core calculation. In the lattice calculation, in which Boltzmann's transport equation is solved over an assembly geometry, the temperature distribution is uniform and constant during irradiation. This represents a set of approximations since, on the one hand, the temperature distribution in the assembly is not uniform (strong temperature gradients in the fuel pins, discrepancies between the fuel pins) and on the other hand, irradiation causes the thermal properties of the pins to change, which modifies the temperature distribution. Our work aims at implementing and introducing a neutronics-thermomechanics coupling into the lattice calculation to finely discretize the temperature distribution and to study its effects. To perform the study, CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) lattice code APOLLO2 was used for neutronics and EDF (Electricite De France) code C3THER was used for the thermal calculations. We show very small effects of the pin-scaled coupling when comparing the use of a temperature profile with the use of an uniform temperature over UOX-type and MOX-type fuels. We next investigate the thermal feedback using an assembly-scaled coupling taking into account the presence of large water gaps on an UOX-type assembly at burnup 0. We show the very small impact on the calculation of the hot spot factor. Finally, the coupling is introduced into the isotopic depletion calculation and we show that reactivity and isotopic number densities deviations remain small albeit not negligible for UOX-type and MOX-type assemblies. The specific behavior of gadolinium-stuffed fuel pins in an UO2Gd2O 3-type assembly is highlighted.

System and method for reducing combustion dynamics in a combustor

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward; Zuo, Baifang; York, William David

2013-08-20

A system for reducing combustion dynamics in a combustor includes an end cap having an upstream surface axially separated from a downstream surface, and tube bundles extend through the end cap. A diluent supply in fluid communication with the end cap provides diluent flow to the end cap. Diluent distributors circumferentially arranged inside at least one tube bundle extend downstream from the downstream surface and provide fluid communication for the diluent flow through the end cap. A method for reducing combustion dynamics in a combustor includes flowing fuel through tube bundles that extend axially through an end cap, flowing a diluent through diluent distributors into a combustion chamber, wherein the diluent distributors are circumferentially arranged inside at least one tube bundle and each diluent distributor extends downstream from the end cap, and forming a diluent barrier in the combustion chamber between at least one pair of adjacent tube bundles.

Experimental investigation on circumferential and axial temperature gradient over fuel channel under LOCA

NASA Astrophysics Data System (ADS)

Yadav, Ashwini Kumar; kumar, Ravi; Gupta, Akhilesh; Chatterjee, Barun; Mukhopadhyay, Deb; Lele, H. G.

2014-06-01

In a nuclear reactor temperature rises drastically in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferential and axial temperature gradients during fully and partially voiding conditions in a fuel channel using 19 pin fuel element simulator. A series of experiments were carried out by supplying power to outer, middle and center rods of 19 pin fuel simulator in ratio of 1.4:1.1:1. The temperature at upper periphery of pressure tube (PT) was slightly higher than at bottom due to increase in local equivalent thermal conductivity from top to bottom of PT. To simulate fully voided conditions PT was pressurized at 2.0 MPa pressure with 17.5 kW power injection. Ballooning initiated from center and then propagates towards the ends and hence axial temperature difference has been observed along the length of PT. For asymmetric heating, upper eight rods of fuel simulator were activated and temperature difference up-to 250 °C has been observed from top to bottom periphery of PT. Such situation creates steep circumferential temperature gradient over PT and could lead to breaching of PT under high pressure.

WWER-1000 core and reflector parameters investigation in the LR-0 reactor

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

Zaritsky, S. M.; Alekseev, N. I.; Bolshagin, S. N.

2006-07-01

Measurements and calculations carried out in the core and reflector of WWER-1000 mock-up are discussed: - the determination of the pin-to-pin power distribution in the core by means of gamma-scanning of fuel pins and pin-to-pin calculations with Monte Carlo code MCU-REA and diffusion codes MOBY-DICK (with WIMS-D4 cell constants preparation) and RADAR - the fast neutron spectra measurements by proton recoil method inside the experimental channel in the core and inside the channel in the baffle, and corresponding calculations in P{sub 3}S{sub 8} approximation of discrete ordinates method with code DORT and BUGLE-96 library - the neutron spectra evaluations (adjustment)more » in the same channels in energy region 0.5 eV-18 MeV based on the activation and solid state track detectors measurements. (authors)« less

Computation of neutron fluxes in clusters of fuel pins arranged in hexagonal assemblies (2D and 3D)

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

Prabha, H.; Marleau, G.

2012-07-01

For computations of fluxes, we have used Carvik's method of collision probabilities. This method requires tracking algorithms. An algorithm to compute tracks (in 2D and 3D) has been developed for seven hexagonal geometries with cluster of fuel pins. This has been implemented in the NXT module of the code DRAGON. The flux distribution in cluster of pins has been computed by using this code. For testing the results, they are compared when possible with the EXCELT module of the code DRAGON. Tracks are plotted in the NXT module by using MATLAB, these plots are also presented here. Results are presentedmore » with increasing number of lines to show the convergence of these results. We have numerically computed volumes, surface areas and the percentage errors in these computations. These results show that 2D results converge faster than 3D results. The accuracy on the computation of fluxes up to second decimal is achieved with fewer lines. (authors)« less

CAPSTONE SENIOR DESIGN - SUPRAMOLECULAR PROTON EXCHANGE MEMBRANES FOR FUEL CELLS

EPA Science Inventory

In order to assume a leading role in the burgeoning hydrogen economy, new infrastructure will be required for fuel cell manufacturing and R&D capabilities. The objective of this proposal is the development of a new generation of advanced proton exchange membrane (PEM) technol...

Comparison of Models for Spacer Grid Pressure Loss in Nuclear Fuel Bundles for One and Two-Phase Flows

NASA Astrophysics Data System (ADS)

Maskal, Alan B.

Spacer grids maintain the structural integrity of the fuel rods within fuel bundles of nuclear power plants. They can also improve flow characteristics within the nuclear reactor core. However, spacer grids add reactor coolant pressure losses, which require estimation and engineering into the design. Several mathematical models and computer codes were developed over decades to predict spacer grid pressure loss. Most models use generalized characteristics, measured by older, less precise equipment. The study of OECD/US-NRC BWR Full-Size Fine Mesh Bundle Tests (BFBT) provides updated and detailed experimental single and two-phase results, using technically advanced flow measurements for a wide range of boundary conditions. This thesis compares the predictions from the mathematical models to the BFBT experimental data by utilizing statistical formulae for accuracy and precision. This thesis also analyzes the effects of BFBT flow characteristics on spacer grids. No single model has been identified as valid for all flow conditions. However, some models' predictions perform better than others within a range of flow conditions, based on the accuracy and precision of the models' predictions. This study also demonstrates that pressure and flow quality have a significant effect on two-phase flow spacer grid models' biases.

The new Section 23 of DO160C/ED14C lightning testing of externally mounted electrical equipment

NASA Astrophysics Data System (ADS)

Burrows, B. J. C.

1991-08-01

The new Section 23 is introduced which has only very recently been fully approved by the RTCA for incorporation into the first revision of DO160C/ED14C. Full threat lightning direct effects testing of equipment is entirely new to DO160, the only existing lightning testing is transient testing for LRU's (Line Replaceable Units) by pin or cable bundle injection methods, for equipment entirely contained within the airframe and assumed to be unaffected by direct effects. This testing required transients of very low amplitude compared with lightning itself, whereas the tests now to be described involve full threat lightning testing, that is using the previously established severe parameters of lightning appropriate to the Zone, such as 200 kA for Zone 1A as in AC20-136. Direct effects (i.e., damage) testing involves normally the lightning current arc attaching to the object under test (or very near to it) so submitting it to full potential for the electric, mechanical, thermal and shock damage which is caused by high current arcing. Since equipment for any part of the airframe require qualification, tests to demonstrate safety of equipment in fuel vapor regions of the airframe are also included.

High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David

2007-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Kapernick, Richard

2007-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

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

Loftus, M J; Hochreiter, L E; McGuire, M F

This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

Vortex Matter in Highly Strained Nb_{75}Zr_{25}: Analogy with Viscous Flow of Disordered Solids

NASA Astrophysics Data System (ADS)

Chandra, Jagdish; Manekar, Meghmalhar; Sharma, V. K.; Mondal, Puspen; Tiwari, Pragya; Roy, S. B.

2017-01-01

We present the results of magnetization and magneto-transport measurements in the superconducting state of an as-cast Nb_{75}Zr_{25} alloy. We also report the microstructure of our sample at various length scales by using optical, scanning electron and transmission electron microscopies. The information of microstructure is used to understand the flux pinning properties in the superconducting state within the framework of collective pinning. The magneto-transport measurements show a non-Arrhenius behaviour of the temperature- and field-dependent resistivity across the resistive transition and is understood in terms of a model for viscous flow of disordered solids which is popularly known as the `shoving model'. The activation energy for flux flow is assumed to be mainly the elastic energy stored in the flux-line lattice. The scaling of pinning force density indicates the presence of two pinning mechanisms of different origins. The elastic constants of the flux-line lattice are used to estimate the length scale of vortex lattice movement, or the volume displaced by the flux-line lattice. It appears that the vortex lattice displacement estimated from elastic energy considerations is of the same order of magnitude as that of the flux bundle hopping length during flux flow. Our results could provide possible directions for establishing a framework where vortex matter and glass-forming liquids or amorphous solids can be treated in a similar manner for understanding the phenomenon of viscous flow in disordered solids or more generally the pinning and depinning properties of elastic manifolds in random media. It is likely that the vortex molasses scenario is more suited to explain the vortex dynamics in conventional low-T_C superconductors.

Fabrication of (U, Zr) C-fueled/tungsten-clad specimens for irradiation in the Plum Brook Reactor Facility

NASA Technical Reports Server (NTRS)

1972-01-01

Fuel samples, 90UC - 10 ZrC, and chemically vapor deposited tungsten fuel cups were fabricated for the study of the long term dimensional stability and compatibility of the carbide-tungsten fuel-cladding systems under irradiation. These fuel samples and fuel cups were assembled into the fuel pins of two capsules, designated as V-2E and V-2F, for irradiation in NASA Plum Brook Reactor Facility at a fission power density of 172 watts/c.c. and a miximum cladding temperature of 1823 K. Fabrication methods and characteristics of the fuel samples and fuel cups prepared are described.

Optimizing the use of a John Deere bundling unit in a southern logging system

Treesearch

Steven Meadows; Tom Gallagher; Dana Mitchell

2009-01-01

With the current energy crisis and with petroleum prices skyrocketing, all sources of alternative fuels need to be explored. John Deere’s Biomass Bundler unit is an effective machine for harvesting forest residues, which can be used as a source of fuel wood and/or a feedstock for bio‐fuel production. This project aims to explore an...

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE PAGES

Piro, M.H.A; Wassermann, F.; Grundmann, S.; ...

2017-05-23

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

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

Piro, M.H.A; Wassermann, F.; Grundmann, S.

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

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.

Heater Development, Fabrication, and Testing: Analysis of Fabricated Heaters

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Dickens, R. E.; Farmer, J. T.; Davis, J. D.; Adams, M. R.; Martin, J. J.; Webster, K. L.

2008-01-01

Thermal simulators (highly designed heater elements) developed at the Early Flight Fission Test Facility (EFF-TF) are used to simulate the heat from nuclear fission in a variety of reactor concepts. When inserted into the reactor geometry, the purpose of the thermal simulators is to deliver thermal power to the test article in the same fashion as if nuclear fuel were present. Considerable effort has been expended to mimic heat from fission as closely as possible. To accurately represent the fuel, the simulators should be capable of matching the overall properties of the nuclear fuel rather than simply matching the fuel temperatures. This includes matching thermal stresses in the pin, pin conductivities, total core power, and core power profile (axial and radial). This Technical Memorandum discusses the historical development of the thermal simulators used in nonnuclear testing at the EFF-TF and provides a basis for the development of the current series of thermal simulators. The status of current heater fabrication and testing is assessed, providing data and analyses for both successes and failures experienced in the heater development and testing program.

An Overview of Reactor Concepts, a Survey of Reactor Designs.

DTIC Science & Technology

1985-02-01

may be very different. HTGRs may use highly enriched uranium, thereby yielding better fuel economy and a reduc- tion of the actual core size for a...specific power level. The HTGR core may have fuel and control rods placed in graphite arrays similar to PWR core con- figuration, or they may have fuel ...rods are pulled out. A Peach Bottom core design is another HTGR design. This design is featured by the fuel pin’s ability to purge itself of fission

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE PAGES

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel--coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

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

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

Fuels irradiation testing for the SP-100 program

NASA Technical Reports Server (NTRS)

Makenas, Bruce J.; Hales, Janell W.; Ward, Alva L.

1991-01-01

An SP-100 fuel pin irradiation testing program is well on the way to providing data for performance correlations and demonstrating the lifetime and safety of the fuel system of the compact lithium-cooled reactor. Key SP-100 fuel performance issues addressed are the need for low fuel swelling and low fission gas release to minimize cladding strain, and the need for barrier integrity to prevent fuel/cladding chemical interaction. This paper provides a description of the irradiation test program that addresses these key issues and summarizes recent results of posttest examinations including data obtained at 6 atom percent goal burnup.

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

DOEpatents

Bradley, John G.

1982-01-01

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

INDEXING MECHANISM

DOEpatents

Kock, L.J.

1959-09-22

A device is presented for loading and unloading fuel elements containing material fissionable by neutrons of thermal energy. The device comprises a combination of mechanical features Including a base, a lever pivotally attached to the base, an Indexing plate on the base parallel to the plane of lever rotation and having a plurality of apertures, the apertures being disposed In rows, each aperture having a keyway, an Index pin movably disposed to the plane of lever rotation and having a plurality of apertures, the apertures being disposed in rows, each aperture having a keyway, an index pin movably disposed on the lever normal to the plane rotation, a key on the pin, a sleeve on the lever spaced from and parallel to the index pin, a pair of pulleys and a cable disposed between them, an open collar rotatably attached to the sleeve and linked to one of the pulleys, a pin extending from the collar, and a bearing movably mounted in the sleeve and having at least two longitudinal grooves in the outside surface.

Critical current density, vortex dynamics, and phase diagram of single-crystal FeSe

NASA Astrophysics Data System (ADS)

Sun, Yue; Pyon, Sunseng; Tamegai, Tsuyoshi; Kobayashi, Ryo; Watashige, Tatsuya; Kasahara, Shigeru; Matsuda, Yuji; Shibauchi, Takasada

2015-10-01

We present a comprehensive study of the vortex pinning and dynamics in a high-quality FeSe single crystal which is free from doping-introduced inhomogeneities and charged quasiparticle scattering because of its innate superconductivity. The critical current density Jc is found to be almost isotropic and reaches a value of ˜3 ×104 A /cm2 at 2 K (self-field) for both H ∥c and a b . The normalized magnetic relaxation rate S (=∣d ln M /d ln t ∣ ) shows a temperature-insensitive plateau behavior in the intermediate temperature range with a relatively high creep rate (S ˜ 0.02 under zero field), which is interpreted in the framework of the collective creep theory. A crossover from the elastic to plastic creep is observed, while the fishtail effect is absent for both H ∥c and a b . Based on this observation, the origin of the fishtail effect is also discussed. Combining the results of Jc and S , the vortex motion in the FeSe single crystal is found to be dominated by sparse, strong pointlike pinning from nanometer-sized defects or imperfections. The weak collective pinning is also observed and proved in the form of large bundles. Besides, the vortex phase diagram of FeSe is also constructed and discussed.

Connections for solid oxide fuel cells

DOEpatents

Collie, Jeffrey C.

1999-01-01

A connection for fuel cell assemblies is disclosed. The connection includes compliant members connected to individual fuel cells and a rigid member connected to the compliant members. Adjacent bundles or modules of fuel cells are connected together by mechanically joining their rigid members. The compliant/rigid connection permits construction of generator fuel cell stacks from basic modular groups of cells of any desired size. The connections can be made prior to installation of the fuel cells in a generator, thereby eliminating the need for in-situ completion of the connections. In addition to allowing pre-fabrication, the compliant/rigid connections also simplify removal and replacement of sections of a generator fuel cell stack.

Nuclear reactor fuel assembly duct-tube-to-handling-socket attachment system

DOEpatents

Christiansen, David W.; Smith, Bob G.

1982-01-01

A reusable system for removably attaching the upper end 10of a nuclear reactor duct tube to the lower end 30 of a nuclear reactor fuel assembly handling socket. A transition ring 20, fixed to the duct tube's upper end 10, has an interior-threaded section 22 with a first locking hole segment 24. An adaptor ring 40, fixed to the handling socket's lower end 30 has an outside-threaded section 42 with a second locking hole segment 44. The inside 22 and outside 42 threaded sections match and can be joined so that the first 24 and second 44 locking hole segments can be aligned to form a locking hole. A locking ring 50, with a locking pin 52, slides over the adaptor ring 40 so that the locking pin 52 fits in the locking hole. A swage lock 60 or a cantilever finger lock 70 is formed from the locking cup collar 26 to fit in a matching groove 54 or 56 in the locking ring 50 to prevent the locking ring's locking pin 52 from backing out of the locking hole.

EXPERIMENTAL STUDIES OF TRANSIENT EFFECTS IN FAST REACTOR FUELS. SERIES I. UO$sub 2$ IRRADIATIONS

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

Field, J.H.

1962-11-15

An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO/sub 2/ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel. (auth)

Study of clad ballooning and rupture behaviour of Indian PHWR fuel pins under transient heating condition in steam environment

NASA Astrophysics Data System (ADS)

Sawarn, Tapan K.; Banerjee, Suparna; Sheelvantra, Smita S.; Singh, J. L.; Bhasin, Vivek

2017-11-01

This paper presents the results of the investigation on the deformation and rupture characteristics of Indian pressurized heavy water reactor (IPHWR) fuel pins under simulated loss of coolant accident (LOCA) condition in steam environment. Transient heating experiments were carried out on single fuel pin internally pressurized with argon gas in the range 3-70 bar. Effect of internal pressure on burst temperature, influence of burst temperature on the circumferential strain and rupture opening area were also studied. Two circumferential strain maxima at the burst temperatures of 740 & ∼979 °C and a minimum at the burst temperature of ∼868 °C were observed. It was found that oxidation had considerable effect on the burst behavior. Test data were used to derive a direct empirical correlation for burst stress exclusively as a function of temperature. The ballooning and rupture behaviours in steam and argon environments have been compared. Experimental data were examined against various correlations using Erbacher equation and author's previous correlation in argon. A second burst correlation has also been developed combining the equation in argon from the previous work of the authors and an exponential factor with oxygen content as a parameter assuming the burst stress to be a function of both temperature and oxygen concentration. The burst temperatures predicted by this empirical correlation are in good agreement with the test data.

Validation of the analytical methods in the LWR code BOXER for gadolinium-loaded fuel pins

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

Paratte, J.M.; Arkuszewski, J.J.; Kamboj, B.K.

1990-01-01

Due to the very high absorption occurring in gadolinium-loaded fuel pins, calculations of lattices with such pins present are a demanding test of the analysis methods in light water reactor (LWR) cell and assembly codes. Considerable effort has, therefore, been devoted to the validation of code methods for gadolinia fuel. The goal of the work reported in this paper is to check the analysis methods in the LWR cell/assembly code BOXER and its associated cross-section processing code ETOBOX, by comparison of BOXER results with those from a very accurate Monte Carlo calculation for a gadolinium benchmark problem. Initial results ofmore » such a comparison have been previously reported. However, the Monte Carlo calculations, done with the MCNP code, were performed at Los Alamos National Laboratory using ENDF/B-V data, while the BOXER calculations were performed at the Paul Scherrer Institute using JEF-1 nuclear data. This difference in the basic nuclear data used for the two calculations, caused by the restricted nature of these evaluated data files, led to associated uncertainties in a comparison of the results for methods validation. In the joint investigations at the Georgia Institute of Technology and PSI, such uncertainty in this comparison was eliminated by using ENDF/B-V data for BOXER calculations at Georgia Tech.« less

Irradiation test of tungsten clad uranium carbide-zirconium carbide ((U,Zr)C) specimens for thermionic reactor application at conditions conductive to long-term performance

NASA Technical Reports Server (NTRS)

Creagh, J. W. R.; Smith, J. R.

1973-01-01

Uranium carbide fueled, thermionic emitter configurations were encapsulated and irradiated. One capsule contained a specimen clad with fluoride derived chemically vapor deposited (CVD) tungsten. The other capsule used a duplex clad specimen consisting of chloride derived on floride derived CVD tungsten. Both fuel pins were 16 millimeters in diameter and contained a 45.7-millimeter length of fuel.

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

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

Todreas, N.E.; Cheng, S.K.; Basehore, K.

1984-08-01

This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized.more » Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.« less

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

76 FR 34770 - Rensselaer Polytechnic Institute Critical Experiments Facility; Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-14

...)) stainless steel tank with an inner diameter of 2.1 m (7 ft). The reactor is fueled with low enriched uranium SPERT fuel pins. Reactivity control is provided by four Boron-10 control rods. A detailed description of... moderator are stainless steel, thus eliminating the need for routine filtration and demineralization of the...

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

Burke, Timothy P.; Martz, Roger L.; Kiedrowski, Brian C.

New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructivemore » Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.« less

Pulsed Magnetic Welding for Advanced Core and Cladding Steel

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

Cao, Guoping; Yang, Yong

2013-12-19

To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pinmore » end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.« less

Shear and friction between carbon nanotubes in bundles and yarns.

PubMed

Paci, Jeffrey T; Furmanchuk, Al'ona; Espinosa, Horacio D; Schatz, George C

2014-11-12

We perform a detailed density functional theory assessment of the factors that determine shear interactions between carbon nanotubes (CNTs) within bundles and in related CNT and graphene structures including yarns, providing an explanation for the shear force measured in recent experiments (Filleter, T. etal. Nano Lett. 2012, 12, 73). The potential energy barriers separating AB stacked structures are found to be irrelevant to the shear analysis for bundles and yarns due to turbostratic stacking, and as a result, the tube-tube shear strength for pristine CNTs is estimated to be <0.24 MPa, that is, extremely small. Instead, it is pinning due to the presence of defects and functional groups at the tube ends that primarily cause resistance to shear when bundles are fractured in weak vacuum (∼10(-5) Torr). Such defects and groups are estimated to involve 0.55 eV interaction energies on average, which is much larger than single-atom vacancy defects (approximately 0.039 eV). Furthermore, because graphitic materials are stiff they have large coherence lengths, and this means that push-pull effects result in force cancellation for vacancy and other defects that are internal to the CNTs. Another important factor is the softness of cantilever structures relative to the stiff CNTs in the experiments, as this contributes to elastic instability transitions that account for significant dissipation during shear that has been observed. The application of these results to the mechanical behavior of yarns is discussed, providing general guidelines for the manufacture of strong yarns composed of CNTs.

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.

76 FR 33991 - Airworthiness Directives; The Boeing Company Model 727, 727C, 727-100, 727-100C, 727-200, and 727...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-10

... manufacturer. We are issuing this AD to increase the level of protection from lightning strikes and prevent the... of protection from lightning strikes and prevent the potential of ignition sources inside fuel tanks... existing unshielded fuel quantity indication system (FQIS) wire bundles with double shielded FQIS wire...

VICTORIA-92 pretest analyses of PHEBUS-FPT0

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

Bixler, N.E.; Erickson, C.M.

FPT0 is the first of six tests that are scheduled to be conducted in an experimental reactor in Cadarache, France. The test apparatus consists of an in-pile fuel bundle, an upper plenum, a hot leg, a steam generator, a cold leg, and a small containment. Thus, the test is integral in the sense that it attempts to simulate all of the processes that would be operative in a severe nuclear accident. In FPT0, the fuel will be trace irradiated; in subsequent tests high burn-up fuel will be used. This report discusses separate pretest analyses of the FPT0 fuel bundle andmore » primary circuit have been conducted using the USNRC`s source term code, VICTORIA-92. Predictions for release of fission product, control rod, and structural elements from the test section are compared with those given by CORSOR-M. In general, the releases predicted by VICTORIA-92 occur earlier than those predicted by CORSOR-M. The other notable difference is that U release is predicted to be on a par with that of the control rod elements; CORSOR-M predicts U release to be about 2 orders of magnitude greater.« less

Hand-held optical fuel pin scanner

DOEpatents

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

1980-12-07

An optical scanner for indicia arranged in a focal plane perpendicular to an optical system including a rotatable dove prism. The dove prism transmits a rotating image to a stationary photodiode array.

Hand-held optical fuel pin scanner

DOEpatents

Kirchner, Tommy L.; Powers, Hurshal G.

1987-01-01

An optical scanner for indicia arranged in a focal plane perpendicular to an optical system including a rotatable dove prism. The dove prism transmits a rotating image to a stationary photodiode array.

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

Ingham, J.G.

The IDENT 1578 container, which is a 110-in. long 5.5-in. OD tube, is designed for shipping FFTF fuel elements in T-3 casks between HEDL, HFEF, and other laboratories. The thermal analysis was conducted to evaluate whether or not the container satisfies its thermal design criteria (handle a decay heat load of 600 watts, max fuel pin cladding temperature not exceeding 800/sup 0/F).

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

DOE PAGES

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

2016-05-05

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

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

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

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

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

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

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

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

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

Preliminary Energy Deposition Calculations for GRIST-2 Tests in the TREAT Upgrade

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

Olson, W. O.

1978-03-01

Preliminary studies have been made to estimate the energy deposition in GRIST-2 tests irradiated in the proposed TREAT Upgrade reactor. The objective of the GRIST-2 project is to test GCFR (gas cooled fast reactor) fuel under conditions of hypothetical core disruptive accidents (HCDA). Test requirements are (1) an energy deposition in the test of approximately 2500 J/g or higher, (2) a pin-to-pin variation in energy deposition of less than 10% and (3) the variation in the energy deposition across any pin (at a given axial position) should be less than 10%. Calculations performed by EG&G Idaho were made for 7more » and 37-pin tests using one-dimensional transport theory. These yield average energy deposition rates in the test at the axial peak which are in the 5000-5500 J/g range for the 37-pin test and are in the 8500-9000 J/g range for the 7-pin test. These values are obtained with a cadmium thermal neutron filter (TNF) surrounding the test. This hardens the flux to meet the third requirement. The central test pin is fully enriched UO{sub 2}, with the outer pins having lower enrichments to satisfy requirement 2. Addition of the TNF reduces the energy deposition by about 10%. The results in the above calculations are also compared with the Monte Carlo results computed by ANL-West personnel.« less

Advanced foil activation techniques for the measurement of within-pin distributions of the 63Cu(n,γ) 64Cu reaction rate in nuclear fuel

NASA Astrophysics Data System (ADS)

Macku, K.; Jatuff, F.; Murphy, M. F.; Joneja, O. P.; Bischofberger, R.; Chawla, R.

2006-06-01

Different foil activation techniques have been used for measuring spatial distributions of the 63Cu(n,γ) 64Cu reaction within two pins of a SVEA-96 Optima2 boiling water reactor fuel assembly, at the critical facility PROTEUS. This reaction is of interest because its 1/v cross-section gives it a good representation of the 235U fission rate. Initially, radial capture rate profiles were measured with mechanically punched copper foils. More detailed profiles were then determined by using a 0.2 mm copper wire spiral (˜200 μm resolution), as well as 5-, 10-, and 20-ring UV-lithography, electroplating, and molding (UV-LIGA) foils (up to a 100 μm resolution). For azimuthal measurements, apart from manually cut activation foils (into 8 sectors), 8- and 12-sector LIGA foils were used. The highly versatile LIGA foils have the additional advantage of being very easily separated into individual pieces after irradiation without the use of punches or other cutting tools. In order to account for the invasive character of the foil activation techniques, corrections to account for sample perturbations and for self-shielding effects were determined via simplified Monte Carlo (MCNP4C) modeling of the experimental setup. The final results from the various measurements of 63Cu(n,γ) 64Cu within-pin distributions have been compared with MCNP computations employing a detailed model of the full SVEA Optima2 fuel assembly.

Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

NASA Astrophysics Data System (ADS)

Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.

2006-06-01

Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

Interconnection of bundled solid oxide fuel cells

DOEpatents

Brown, Michael; Bessette, II, Norman F; Litka, Anthony F; Schmidt, Douglas S

2014-01-14

A system and method for electrically interconnecting a plurality of fuel cells to provide dense packing of the fuel cells. Each one of the plurality of fuel cells has a plurality of discrete electrical connection points along an outer surface. Electrical connections are made directly between the discrete electrical connection points of adjacent fuel cells so that the fuel cells can be packed more densely. Fuel cells have at least one outer electrode and at least one discrete interconnection to an inner electrode, wherein the outer electrode is one of a cathode and and anode and wherein the inner electrode is the other of the cathode and the anode. In tubular solid oxide fuel cells the discrete electrical connection points are spaced along the length of the fuel cell.

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.

A strategy for intensive production of molybdenum-99 isotopes for nuclear medicine using CANDU reactors.

PubMed

Morreale, A C; Novog, D R; Luxat, J C

2012-01-01

Technetium-99m is an important medical isotope utilized worldwide in nuclear medicine and is produced from the decay of its parent isotope, molybdenum-99. The online fueling capability and compact fuel of the CANDU(®)(1) reactor allows for the potential production of large quantities of (99)Mo. This paper proposes (99)Mo production strategies using modified target fuel bundles loaded into CANDU fuel channels. Using a small group of channels a yield of 89-113% of the weekly world demand for (99)Mo can be obtained. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

Ketusky, E.

The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtainedmore » individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.« less

Space and Time Distribution of Pu Isotopes inside The First Experimental Fuel Pin Designed for PWR and Manufactured in Indonesia

NASA Astrophysics Data System (ADS)

Suwardi; Setiawan, J.; Susilo, J.

2017-01-01

The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared and planned to be tested in power ramp irradiation. An irradiation test should be designed to allow an experiment can be performed safely and giving maximum results of many performance aspects of design and manufacturing. Performance analysis to the fuel specimen shows that the specimen is not match to be used for power ramp testing. Enlargement by 0.20 mm of pellet diameter has been proposed. The present work is evaluation of modified design for important aspect of isotopic Pu distribution during irradiation test, because generated Pu isotopes in natural UO2 fuel, contribute more power relative to the contribution by enriched UO2 fuel. The axial profile of neutrons flux have been chosen from both experimental measurement and model calculation. The parameters of ramp power has been obtained from statistical experiment data. A simplified and typical base-load commercial PHWR profile of LHR history has been chosen, to determine the minimum irradiation time before ramp test can be performed. The data design and Mat pro XI materials properties models have been chosen. The axial profile of neutrons flux has been accommodated by 5 slices of discrete pin. The Pu distribution of slice-4 with highest power rate has been chosen to be evaluated. The radial discretion of pellet and cladding and numerical parameter have been used the default best practice of TU. The results shows that Pu 239 increased rapidly. The maximum burn up of slice 4 at upper the median slice, it reached nearly 90% of maximum value at about 6000 h with peak of 0.8%a Pu/HM at 22000 h, which is higher than initial U 235. Each 240, 241 and 240 Pu grows slower and ends up to 0.4, 0.2 and 0.18 % respectively. This results can be used for verification of other aspect of fuel behavior in the modeling results and also can be used as guide and comparison to the future post irradiation examination for Pu isotopes distribution.

EVALUATION OF DATA OBTAINED ON "MANUFACTURING PROCESS" DEVELOPMENT BUNDLES PD 1 THROUGH 5 PRIOR TO MACHINING OPERATIONS

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

Frankhouser, W.L.; Eyler, J.H.

1956-07-24

Five reference fuel rod bundles were welded and evaluated dimensionally. Dimensional data are presented for the as-welded condition and for the annealed bundle with spacer strips removed (prior to the final machining operations). The welding sequence developed for Core Manufacturing should provide A'' boundles in respect to rod spacing measurements. It will probably not be possible to meet the same requirements for water channel averages, because the design tolerances are not consistent with some factors inherent to the production process. A method to improve this situation is presented. The data presented were evaluated in a fashion similar to that whichmore » would be used in the proposed scheme. Rods tended to bow resulting in a slightly barrel-shaped'' boundle. It is believed this condition can be overcome by providing special bundle peripheral clamps during annealing. Rod distortion should also be reduced by a redesign and relocation of strip spacers. The new design is proposed. (auth)« less

Design of a highly parallel board-level-interconnection with 320 Gbps capacity

NASA Astrophysics Data System (ADS)

Lohmann, U.; Jahns, J.; Limmer, S.; Fey, D.; Bauer, H.

2012-01-01

A parallel board-level interconnection design is presented consisting of 32 channels, each operating at 10 Gbps. The hardware uses available optoelectronic components (VCSEL, TIA, pin-diodes) and a combination of planarintegrated free-space optics, fiber-bundles and available MEMS-components, like the DMD™ from Texas Instruments. As a specific feature, we present a new modular inter-board interconnect, realized by 3D fiber-matrix connectors. The performance of the interconnect is evaluated with regard to optical properties and power consumption. Finally, we discuss the application of the interconnect for strongly distributed system architectures, as, for example, in high performance embedded computing systems and data centers.

Optical Fibre Sensor For Measuring pH In Physiological Range

NASA Astrophysics Data System (ADS)

Golunski, Witold; Hypszer, Ryszard; Plucinski, Jerzy

1990-01-01

The principle of fibre optic pH sensor operation is given in this paper. PH measurement in 7.0-7.5 range is based on changing of optical property of a indicator. The indicator is sensitive to the hydrogen ion concentration in the water solution. Microspheres of the polymer XAD-2 (a styrene-divinylbenzene copolymer) containing bound phenol red were used as a indicator. Such prepared indicator was inserted in optrode. The optrode was connected with transmitter and receiver by a bundle of glass fibres (multicomponent glass). Transmitter was done by using green LED while receiver construction was based on pin photodiode.

Light Water Breeder Reactor fuel rod design and performance characteristics (LWBR Development Program)

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

Campbell, W.R.; Giovengo, J.F.

1987-10-01

Light Water Breeder Reactor (LWBR) fuel rods were designed to provide a reliable fuel system utilizing thorium/uranium-233 mixed-oxide fuel while simultaneously minimizing structural material to enhance fuel breeding. The fuel system was designed to be capable of operating successfully under both load follow and base load conditions. The breeding objective required thin-walled, low hafnium content Zircaloy cladding, tightly spaced fuel rods with a minimum number of support grid levels, and movable fuel rod bundles to supplant control rods. Specific fuel rod design considerations and their effects on performance capability are described. Successful completion of power operations to over 160 percentmore » of design lifetime including over 200 daily load follow cycles has proven the performance capability of the fuel system. 68 refs., 19 figs., 44 tabs.« less

Cap assembly for a bundled tube fuel injector

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

LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold

2016-04-26

A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate.more » A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.« less

Application of the SHARP Toolkit to Sodium-Cooled Fast Reactor Challenge Problems

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

Shemon, E. R.; Yu, Y.; Kim, T. K.

The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit is under development by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of the U.S. Department of Energy, Office of Nuclear Energy. To better understand and exploit the benefits of advanced modeling simulations, the NEAMS Campaign initiated the “Sodium-Cooled Fast Reactor (SFR) Challenge Problems” task, which include the assessment of hot channel factors (HCFs) and the demonstration of zooming capability using the SHARP toolkit. If both challenge problems are resolved through advanced modeling and simulation using the SHARP toolkit, the economic competitiveness of a SFR can be significantly improved. The effortsmore » in the first year of this project focused on the development of computational models, meshes, and coupling procedures for multi-physics calculations using the neutronics (PROTEUS) and thermal-hydraulic (Nek5000) components of the SHARP toolkit, as well as demonstration of the HCF calculation capability for the 100 MWe Advanced Fast Reactor (AFR-100) design. Testing the feasibility of the SHARP zooming capability is planned in FY 2018. The HCFs developed for the earlier SFRs (FFTF, CRBR, and EBR-II) were reviewed, and a subset of these were identified as potential candidates for reduction or elimination through high-fidelity simulations. A one-way offline coupling method was used to evaluate the HCFs where the neutronics solver PROTEUS computes the power profile based on an assumed temperature, and the computational fluid dynamics solver Nek5000 evaluates the peak temperatures using the neutronics power profile. If the initial temperature profile used in the neutronics calculation is reasonably accurate, the one-way offline method is valid because the neutronics power profile has weak dependence on small temperature variation. In order to get more precise results, the proper temperature profile for initial neutronics calculations was obtained from the STAR-CCM+ calculations. The HCFs of the peak temperatures at cladding outer surface, cladding inner wall surface, and fuel centerline induced by cladding manufacturing tolerance and uncertainties on the cladding, coolant, and fuel properties were evaluated for the AFR-100. Some assessment on the effect of wire wrap configuration and size of the bundle shows that it is practical to use the 7-pin bare rod bundle to calculate the HCFs. The resulting HCFs obtained from the high-fidelity SHARP calculations are generally smaller than those developed for the earlier SFRs because the most uncertainties involved in the modeling and simulations were disappeared. For completeness, additional investigations are planned in FY 2018, which will use random sampling techniques.« less

MOOSE Implementation of MAMBA

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

Galloway, Jack; Matthews, Topher

The development of MAMBA is targeted at capturing both core wide CRUD induced power shifts (CIPS) as well as pin-­level CRUD induced localized corrosion (CILC). Both CIPS and CILC require some sort of information from thermal-­hydraulic, neutronics, and fuel performance codes, although the degree of coupling is different for the two effects. Since CIPS necessarily requires a core-­wide power distribution solve, it requires tight coupling with a neutronics code. Conversely, CIPS tends to be an individual pin phenomenon, requiring tight coupling a fuel performance code. As efforts are now focused on coupling MAMBA within the VERA suite, a natural separationmore » has surfaced in which a FORTRAN rewrite of MAMBA is optimal for VERA integration to capture CIPS behavior, while a CILC focused calculation would benefit from a tight coupling with BISON, motivating a MOOSE version of MAMBA.« less

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

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.

Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

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

Jason M. Harp; Paul A. Demkowicz

2014-10-01

In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10 -4 to 10 -5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materialsmore » is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.« less

Irradiation and post-irradiation examination of uranium-free nitride fuel

NASA Astrophysics Data System (ADS)

Hania, P. R.; Klaassen, F. C.; Wernli, B.; Streit, M.; Restani, R.; Ingold, F.; Fedorov, A. V.; Wallenius, J.

2015-11-01

Two identical Phénix-type 15-15Ti steel pinlets each containing a 70 mm Pu0.3Zr0.7N fuel stack in a 1-bar helium atmosphere have been irradiated in the HFR Petten at medium high linear power (46-47 kW/m at BOL) and an average cladding temperature of 505 °C. The pins were irradiated to a plutonium burn-up of 9.7% (88 MWd/kgHM) in 170 full power days. Both pins remained fully intact. Post-irradiation examination performed at NRG and PSI showed that the overall swelling rate of the fuel was 0.92 vol-%/%FIHMA. Fission gas release was 5-6%, while helium release was larger than 50%. No fuel restructuring was observed, and only mild cracking. EPMA measurements show a burn-up increase toward the pellet edge of up to 4 times. All investigated fission products except to some extent the noble metals were found to be evenly distributed over the matrix, indicating good solubility. Local formation of a secondary phase with high Pu content and hardly any Zr was observed. A general conclusion of this investigation is that ZrN is a suitable inert matrix for burning plutonium at high destruction rates.

Uranium dioxide fuel cladding strain investigation with the use of CYGRO-2 computer program

NASA Technical Reports Server (NTRS)

Smith, J. R.

1973-01-01

Previously irradiated UO2 thermionic fuel pins in which gross fuel-cladding strain occurred were modeled with the use of a computer program to define controlling parameters which may contribute to cladding strain. The computed strain was compared with measured strain, and the computer input data were studied in an attempt to get agreement with measured strain. Because of the limitations of the program and uncertainties in input data, good agreement with measured cladding strain was not attained. A discussion of these limitations is presented.

Uranium nitride fuel fabrication for SP-100 reactors

NASA Technical Reports Server (NTRS)

Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

1987-01-01

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

Uranium nitride fuel fabrication for SP-100 reactors

NASA Astrophysics Data System (ADS)

Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

Documentation of Stainless Steel Lithium Circuit Test Section Design. Suppl

NASA Technical Reports Server (NTRS)

Godfroy, Thomas J. (Compiler); Martin, James J.

2010-01-01

The Early Flight Fission-Test Facilities (EFF-TF) team was tasked by Naval Reactors Prime Contract Team (NRPCT) to design, fabricate, and test an actively pumped lithium (Li) flow circuit. This Li circuit takes advantage of work in progress at the EFF TF on a stainless steel sodium/potassium (NaK) circuit. The effort involved modifying the original stainless steel NaK circuit such that it could be operated with Li in place of NaK. This new design considered freeze/thaw issues and required the addition of an expansion tank and expansion/extrusion volumes in the circuit plumbing. Instrumentation has been specified for Li and circuit heaters have been placed throughout the design to ensure adequate operational temperatures and no uncontrolled freezing of the Li. All major components have been designed and fabricated prior to circuit redesign for Li and were not modified. Basic circuit components include: reactor segment, Li to gas heat exchanger, electromagnetic liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. The reactor segment, based on a Los Alamos National Laboratory 100-kW design study with 120 fuel pins, is the only prototypic component in the circuit. However, due to earlier funding constraints, a 37-pin partial-array of the core, including the central three rings of fuel pins (pin and flow path dimensions are the same as those in the full design), was selected for fabrication and test. This Technical Publication summarizes the design and integration of the pumped liquid metal Li flow circuit as of May 1, 2005. This supplement contains drawings, analysis, and calculations

Documentation of Stainless Steel Lithium Circuit Test Section Design

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Martin, J. J.; Stewart, E. T.; Rhys, N. O.

2010-01-01

The Early Flight Fission-Test Facilities (EFF-TF) team was tasked by Naval Reactors Prime Contract Team (NRPCT) to design, fabricate, and test an actively pumped lithium (Li) flow circuit. This Li circuit takes advantage of work in progress at the EFF TF on a stainless steel sodium/potassium (NaK) circuit. The effort involved modifying the original stainless steel NaK circuit such that it could be operated with Li in place of NaK. This new design considered freeze/thaw issues and required the addition of an expansion tank and expansion/extrusion volumes in the circuit plumbing. Instrumentation has been specified for Li and circuit heaters have been placed throughout the design to ensure adequate operational temperatures and no uncontrolled freezing of the Li. All major components have been designed and fabricated prior to circuit redesign for Li and were not modified. Basic circuit components include: reactor segment, Li to gas heat exchanger, electromagnetic liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. The reactor segment, based on a Los Alamos National Laboratory 100-kW design study with 120 fuel pins, is the only prototypic component in the circuit. However, due to earlier funding constraints, a 37-pin partial-array of the core, including the central three rings of fuel pins (pin and flow path dimensions are the same as those in the full design), was selected for fabrication and test. This Technical Publication summarizes the design and integration of the pumped liquid metal Li flow circuit as of May 1, 2005.

Premixed direct injection nozzle for highly reactive fuels

DOEpatents

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

2013-09-24

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

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

Becker, Ines; Schillig, Cora

A double-sided adhesive metal-based tape for use as contacting aid for SOFC fuel cells is provided. The double-sided metal-based adhesive tape is suitable for simplifying the construction of cell bundles. The double-sided metal-based adhesive tape is used for electrical contacting of the cell connector with the anode and for electrical contacting of the interconnector of the fuel cells with the cell connector. A method for producing the double-sided adhesive metal-base tape is also provided.

COMPARISON OF PRESSURE DROP PRODUCED BY SPIRAL WRAPS, COOKIE CUTTERS AND OTHER ROD BUNDLE SPACERS

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

Noyes, R.C.

The problem of predicting pressure drop due to various fuel bundle spacers is considered in some detail. Three sets of experimental data are reviewed and presented in reduced form. These data are compared to several semitheoretical approaches to pressure drop prediction and a best method is selected to make the required predictions. The comparison between predictions for the ASCR spiral wrap spacer and cookie cutter spacer shows that both types of spacers produce about the same pressure drop. Spacer pressure drop is shown to be strongly dependent on spacer frontal area and pitch. (auth)

Enrichment Mechanism of Semiconducting Single-walled Carbon Nanotubes by Surfactant Amines

PubMed Central

Ju, Sang-Yong; Utz, Marcel; Papadimitrakopoulos, Fotios

2009-01-01

Utilization of single-walled carbon nanotubes (SWNTs) in high-end applications hinges on separating metallic (met-) from semiconducting (sem-) SWNTs. Surfactant amines, like octadecylamine (ODA) have proven instrumental for the selective extraction of sem-SWNTs from tetrahydrofuran (THF) nanotube suspensions. The chemical shift differences along the tail of an asymmetric, diacetylenic surfactant amine were used to probe the molecular dynamics in the presence and absence of nanotubes via NMR. The results suggest that the surfactant amine head is firmly immobilized onto the nanotube surface together with acidic water, while the aliphatic tail progressively gains larger mobility as it gets farther from the SWNT. X-ray and high-resolution TEM studies indicate that the sem-enriched sample is populated mainly by small nanotube bundles containing ca. three SWNTs. Molecular simulations in conjunction with previously determined HNO3/H2SO4 oxidation depths for met- and sem-SWNTs indicate that the strong pinning of the amine surfactants on the sem-enriched SWNTs bundles is a result of a well-ordered arrangement of nitrate/amine salts separated with a monomolecular layer of H2O. Such continuous 2D arrangement of nitrate/amine salts shields the local environment adjacent to sem-enriched SWNTs bundles and maintains an acidic pH that preserves nanotube oxidation (i.e. SWNTn+). This, in turn, results in strong interactions with charge-balancing NO3- counter ions that through their association with neutralized surfactant amines provide effective THF dispersion and consequent sem-enrichment. PMID:19397291

Optical fuel pin scanner

DOEpatents

Kirchner, Tommy L.; Powers, Hurshal G.

1983-01-01

An optical scanner for indicia arranged in a focal plane at a cylindrical outside surface by use of an optical system including a rotatable dove prism. The dove prism transmits a rotating image of an encircled cylindrical surface area to a stationary photodiode array.

Nuclear reactor reflector

DOEpatents

Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

1994-01-01

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Nuclear reactor reflector

DOEpatents

Hopkins, R.J.; Land, J.T.; Misvel, M.C.

1994-06-07

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

Distributed temperature sensing inside a 19-rod bundle

DOE PAGES

Lomperski, S.; Bremer, N.; Gerardi, C.

2017-05-23

The temperature field within a model of a sodium-cooled fast reactor fuel rod bundle was measured using Ø155 μm fiber optic distributed temperature sensors (DTS). The bundle consists of 19 electrically-heated rods Ø6.3 mm and 865 mm long. Working fluids were argon and air at atmospheric pressure and Reynolds numbers up to 300. A 20 m-long DTS was threaded through Ø1 mm capillaries wound around rods as wire-wraps. The sensor generated 173 measurements along each rod at 5 mm resolution for a total of 3300 data locations. A second DTS, 58 m long, was suspended between rods to provide 9300more » fluid temperature measurements at 20 mm resolution. Such data density makes it possible to construct 3D maps of the temperature field that are beyond the reach of traditional sensors such as thermocouples. This is illustrated through a series of steady-state and transient tests. As a result, the work demonstrates the feasibility of mapping temperature within the close confines of a rod bundle at resolutions suitable for validation of computational fluid dynamics codes.« less

Experimental evaluation of the sensitivity to fuel utilization and air management on a 100 kW SOFC system

NASA Astrophysics Data System (ADS)

Santarelli, M.; Leone, P.; Calì, M.; Orsello, G.

The tubular SOFC generator CHP-100, built by Siemens Power Generation (SPG) Stationary Fuel Cells (SFC), is running at the Gas Turbine Technologies (GTT) in Torino (Italy), in the framework of the EOS Project. The nominal load of the generator ensures a produced electric power of around 105 kW e ac and around 60 kW t of thermal power at 250 °C to be used for the custom tailored HVAC system. Several experimental sessions have been scheduled on the generator; the aim is to characterize the operation through the analysis of some global performance index and the detailed control of the operation of the different bundles of the whole stack. All the scheduled tests have been performed by applying the methodology of design of experiment; the main obtained results show the effect of the change of the analysed operating factors in terms of distribution of voltage and temperature over the stack. Fuel consumption tests give information about the sensitivity of the voltage and temperature distribution along the single bundles. On the other hand, since the generator is an air cooled system, the results of the tests on the air stoichs have been used to analyze the generator thermal management (temperature distribution and profiles) and its effect on the polarization. The sensitivity analysis of the local voltage to the overall fuel consumption modifications can be used as a powerful procedure to deduce the local distribution of fuel utilization (FU) along the single bundles: in fact, through a model obtained by deriving the polarization curve respect to FU, it is possible to link the distribution of voltage sensitivities to FC to the distribution of the local FU. The FU distribution will be shown as non-uniform, and this affects the local voltage and temperatures, causing a high warming effect in some rows of the generator. Therefore, a discussion around the effectiveness of the thermal regulation made by the air stoichs, in order to reduce the non-uniform distribution of temperature and the overheating (increasing therefore the voltage behavior along the generator) has been performed. It is demonstrated that the utilization of one air plenum is not effective in the thermal regulation of the whole generator, in particular in the reduction of the temperature gradients linked to the non-uniform fuel distribution.

NASA Tech Briefs, February 2007

NASA Technical Reports Server (NTRS)

2007-01-01

Topics covered include: Calibration Test Set for a Phase-Comparison Digital Tracker; Wireless Acoustic Measurement System; Spiral Orbit Tribometer; Arrays of Miniature Microphones for Aeroacoustic Testing; Predicting Rocket or Jet Noise in Real Time; Computational Workbench for Multibody Dynamics; High-Power, High-Efficiency Ka-Band Space Traveling-Wave Tube; Gratings and Random Reflectors for Near-Infrared PIN Diodes; Optically Transparent Split-Ring Antennas for 1 to 10 GHz; Ice-Penetrating Robot for Scientific Exploration; Power-Amplifier Module for 145 to 165 GHz; Aerial Videography From Locally Launched Rockets; SiC Multi-Chip Power Modules as Power-System Building Blocks; Automated Design of Restraint Layer of an Inflatable Vessel; TMS for Instantiating a Knowledge Base With Incomplete Data; Simulating Flights of Future Launch Vehicles and Spacecraft; Control Code for Bearingless Switched- Reluctance Motor; Machine Aided Indexing and the NASA Thesaurus; Arbitrating Control of Control and Display Units; Web-Based Software for Managing Research; Driver Code for Adaptive Optics; Ceramic Paste for Patching High-Temperature Insulation; Fabrication of Polyimide-Matrix/Carbon and Boron-Fiber Tape; Protective Skins for Aerogel Monoliths; Code Assesses Risks Posed by Meteoroids and Orbital Debris; Asymmetric Bulkheads for Cylindrical Pressure Vessels; Self-Regulating Water-Separator System for Fuel Cells; Self-Advancing Step-Tap Drills; Array of Bolometers for Submillimeter- Wavelength Operation; Delta-Doped CCDs as Detector Arrays in Mass Spectrometers; Arrays of Bundles of Carbon Nanotubes as Field Emitters; Staggering Inflation To Stabilize Attitude of a Solar Sail; and Bare Conductive Tether for Decelerating a Spacecraft.

Analysis of the ORNL/TSF GCFR Grid-Plate Shield Design Confirmation Experiment

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

Slater, C.O.; Cramer, S.N.; Ingersoll, D.T.

1979-08-01

The results of the analysis of the GCFR Grid-Plate Shield Design Confirmation Experiment are presented. The experiment, performed at the ORNL Tower Shielding Facility, was designed to test the adequacy of methods and data used in the analysis of the GCFR design. In particular, the experiment tested the adequacy of methods to calculate: (1) axial neutron streaming in the GCFR core and axial blanket, (2) the amount and location of the maximum fast-neutron exposure to the grid plate, and (3) the neutron source leaving the top of the grid plate and entering the upper plenum. Other objectives of the experimentmore » were to verify the grid-plate shielding effectiveness and to assess the effects of fuel-pin and subassembly spacing on radiation levels in the GCFR. The experimental mockups contained regions representing the GCFR core/blanket region, the grid-plate shield section, and the grid plate. Most core design options were covered by allowing: (1) three different spacings between fuel subassemblies, (2) two different void fractions within a subassembly by variation of the number of fuel pins, and (3) a mockup of a control-rod channel.« less

The Pitt Innovation Challenge (PInCh): Driving Innovation in Translational Research Through an Incentive-Based, Problem-Focused Competition.

PubMed

Fitzpatrick, Nicole Edgar; Maier, John; Yasko, Laurel; Mathias, David; Qua, Kacy; Wagner, Erika; Miller, Elizabeth; Reis, Steven E

2017-05-01

Translational research aims to move scientific discoveries across the biomedical spectrum from the laboratory to humans, and to ultimately transform clinical practice and public health policies. Despite efforts to accelerate translational research through national initiatives, several major hurdles remain. The authors created the Pitt Innovation Challenge (PInCh) as an incentive-based, problem-focused approach to solving identified clinical or public health problems at the University of Pittsburgh Clinical and Translational Science Institute in spring 2014. With input from a broad range of stakeholders, PInCh leadership arrived at the challenge question: How do we empower individuals to take control of their own health outcomes? The authors developed the PInCh's three-round proposal submission and review process as well as an online contest management tool to support the process. Ninety-two teams submitted video proposals in round one. Proposals included mobile applications (29; 32%), other information technology (19; 21%), and community program (22; 24%) solutions. Ten teams advanced to the final round, where three were awarded $100,000 to implement their solution over 12 months. In a 6-month follow-up survey, 6/11 (55%) team leaders stated the PInCh helped to facilitate connections outside their normal sphere of collaborators. Additional educational training sessions related to problem-focused research will be developed. The PInCh will be expanded to engage investment and industry communities to facilitate the translation of solutions to clinical practice via commercialization pathways. External organizations and other universities will be engaged to use the PInCh as a mechanism to fuel innovation in their spaces.

Vortex Flux Pinning in Type-Ii Superconductors

NASA Astrophysics Data System (ADS)

Hasan, Mohammad-Khair A. M.

1995-01-01

Rotational magnetization vector measurements on polycrystalline samples of rm YBa_2Cu _3O_7 (YBCO) and (Ba, K)BiO _3 at various fixed fields (H) and temperatures (T) reveal that the vortex flux density (B) in a rotational state consists of a component B_{rm R}, which rotates rigidly with sample rotation, and a B_{rm F} component, which stays at a fixed frictional angle (theta _{rm F}) relative to H. Also, B_{rm R} decreases and ultimately vanishes with increasing H, while B _{rm F} grows monotonically, implying that the vortex pinning strength have a broad distribution. This has been confirmed by the measurements on YBCO of the remanent flux density B^ {rm rm} which can be decomposed analogously into B_{R} ^{} and B_ {F}^{} at angle theta_{F}^{} relative to H. The quantity Hsin theta_{rm F},, which at equilibrium equals tau_{rm p}/mu (the average pinning torque per vortex of moment mu) decreases with increasing high H. This result and the distribution in the strength of the pinning are shown to be consistent with the collective pinning process of vortex bundling. At fixed H, tau_{rm p} decreases rapidly with increasing T, varying approximately as T^{-0.8} for both samples. For polycrystalline YBCO at 4.2 K, B_ {rm R} and B_{ rm F} are found to relax differently with time. The negative creep sign of B_ {rm R} indicates that the number of rotational vortices decreases with time, whereas B _{rm F} shows a positive creep with a negative change in theta_ {rm F}, which indicates that more frictional vortices enter the sample with a tendency of alignment in the direction of H. For grain-oriented YBCO at 4.2 K, the vortex creep measurements of B along the c-axis at different fields showed that: whenever the hysteretic changes of H are reversed in sign, the vortex flux creep (dB/dlogt) decreases very rapidly to zero, where it lingers before changing sign. At the same turning values of H, (dB/dH) also goes to zero. These properties are attributable to the reversals of the vortex motion which occur at the turning values of H and cause a reversal of frictional pinning forces.

Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

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

Davis, W. Jr.; West, G.A.; Stacy, R.G.

1979-03-22

Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO/sub 2/ or 96 to 97% ThO/sub 2/--3 to 4% UO/sub 2/. Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO/sub 2/ or ThO/sub 2/--UO/sub 2/ sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared intomore » lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO/sub 2/ from BWRs and of Zircaloy-4-clad UO/sub 2/ from PWRs. Median particle sizes of UO/sub 2/ from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 ..mu..m; particle sizes of ThO/sub 2/--UO/sub 2/, under these same conditions, ranged from 137 to 202 ..mu..m. Similarly, median particle sizes of UO/sub 2/ from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 ..mu..m. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels; however, unirradiated fuel from vendors was not available for performing comparative shearing experiments. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution estimates can be made of fractions of dislodged fuel having dimensions less than specified values.« less

Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

NASA Astrophysics Data System (ADS)

Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

2016-04-01

This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

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

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

NASA Astrophysics Data System (ADS)

Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

2014-06-01

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

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

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

Material tradeoffs to minimize flooding reactivity

NASA Technical Reports Server (NTRS)

Protsik, R.; Cowan, C. L.; Kangilaski, M.; Vaidyanathan, S.

1987-01-01

In the GE 100 KWe SP-100 design, the array of fuel pins in the core was surrounded by a thin neutron barrier made of a mixture of Nb-1 percent Zr and 20 percent B4C. The purpose of this barrier was to prevent the return to the reactor of thermalized neutrons in the event of a reactor flooding accident. Neutronically, the barrier produced a reduction in k sub eff of 15 percent. However, a mixture of Nb-1 percent Zr and B4C is a development item and may not be possible to produce. Several alternatives to this design were investigated. These include the use of W-26 percent Re for the fuel barrier and the introduction of fuel pin ducts made of Ta-8 percent W-2 percent Hf, Hf-10 percent Nb or pure Re. The use of W-26 percent Re for the fuel barrier material in place of W, and the use of T111, Hf-10 percent Nb, or Re duct material in place of Nb-1 percent Zr or Nb-1 percent Zr-20 percent B4C were found to substantially decrease k sub eff in the flooding situation. The best choice neutronically for duct material is Hf-10 percent Nb, because of its small impact on the k sub eff of the normal core.

Dish/stirling hybrid-receiver

DOEpatents

Mehos, Mark S.; Anselmo, Kenneth M.; Moreno, James B.; Andraka, Charles E.; Rawlinson, K. Scott; Corey, John; Bohn, Mark S.

2002-01-01

A hybrid high-temperature solar receiver is provided which comprises a solar heat-pipe-receiver including a front dome having a solar absorber surface for receiving concentrated solar energy, a heat pipe wick, a rear dome, a sidewall joining the front and the rear dome, and a vapor and a return liquid tube connecting to an engine, and a fossil fuel fired combustion system in radial integration with the sidewall for simultaneous operation with the solar heat pipe receiver, the combustion system comprising an air and fuel pre-mixer, an outer cooling jacket for tangentially introducing and cooling the mixture, a recuperator for preheating the mixture, a burner plenum having an inner and an outer wall, a porous cylindrical metal matrix burner firing radially inward facing a sodium vapor sink, the mixture ignited downstream of the matrix forming combustion products, an exhaust plenum, a fossil-fuel heat-input surface having an outer surface covered with a pin-fin array, the combustion products flowing through the array to give up additional heat to the receiver, and an inner surface covered with an extension of the heat-pipe wick, a pin-fin shroud sealed to the burner and exhaust plenums, an end seal, a flue-gas diversion tube and a flue-gas valve for use at off-design conditions to limit the temperature of the pre-heated air and fuel mixture, preventing pre-ignition.

Mechanisms Underlying the Active Self-Assembly of Microtubule Rings and Spools.

PubMed

VanDelinder, Virginia; Brener, Stephanie; Bachand, George D

2016-03-14

Active self-assembly offers a powerful route for the creation of dynamic multiscale structures that are presently inaccessible with standard microfabrication techniques. One such system uses the translation of microtubule filaments by surface-tethered kinesin to actively assemble nanocomposites with bundle, ring, and spool morphologies. Attempts to observe mechanisms involved in this active assembly system have been hampered by experimental difficulties with performing observation during buffer exchange and photodamage from fluorescent excitation. In the present work, we used a custom microfluidic device to remove these limitations and directly study ring/spool formation, including the earliest events (nucleation) that drive subsequent nanocomposite assembly. Three distinct formation events were observed: pinning, collisions, and induced curvature. Of these three, collisions accounted for the majority of event leading to ring/spool formation, while the rate of pinning was shown to be dependent on the amount of photodamage in the system. We further showed that formation mechanism directly affects the diameter and rotation direction of the resultant rings and spools. Overall, the fundamental understanding described in this work provides a foundation by which the properties of motor-driven, actively assembled nanocomposites may be tailored toward specific applications.

Mechanisms underlying the active self-assembly of microtubule rings and spools

DOE PAGES

VanDelinder, Virginia; Brener, Stephanie; Bachand, George D.

2016-02-04

Here, active self-assembly offers a powerful route for the creation of dynamic multiscale structures that are presently inaccessible with standard microfabrication techniques. One such system uses the translation of microtubule filaments by surface-tethered kinesin to actively assemble nanocomposites with bundle, ring, and spool morphologies. Attempts to observe mechanisms involved in this active assembly system have been hampered by experimental difficulties with performing observation during buffer exchange and photodamage from fluorescent excitation. In the present work, we used a custom microfluidic device to remove these limitations and directly study ring/spool formation, including the earliest events (nucleation) that drive subsequent nanocomposite assembly.more » Three distinct formation events were observed: pinning, collisions, and induced curvature. Of these three, collisions accounted for the majority of event leading to ring/spool formation, while the rate of pinning was shown to be dependent on the amount of photodamage in the system. We further showed that formation mechanism directly affects the diameter and rotation direction of the resultant rings and spools. Overall, the fundamental understanding described in this work provides a foundation by which the properties of motor-driven, actively assembled nanocomposites may be tailored toward specific applications.« less

Gas tungsten arc welder

DOEpatents

Christiansen, D.W.; Brown, W.F.

A welder for automated closure of fuel pins by a gas tungsten arc process in which a rotating length of cladding is positioned adjacent a welding electrode in a sealed enclosure. An independently movable axial grinder is provided in the enclosure for refurbishing the used electrode between welds.

Accelerator-Driven Subcritical System for Disposing of the U.S. Spent Nuclear Fuel Inventory

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

Gohar, Yousry; Cao, Yan; Kraus, Adam R.

The current United States inventory of the spent nuclear fuel (SNF) is ~80,000 metric tons of heavy metal (MTHM), including ~131 tons of minor actinides (MAs) and ~669 tons of plutonium. This study describes a conceptual design of an accelerator-driven subcritical (ADS) system for disposing of this SNF inventory by utilizing the 131 tons of MAs inventory and a fraction of the plutonium inventory for energy production, and transmuting some long-lived fission products. An ADS system with a homogeneous subcritical fission blanket was first examined. A spallation neutron source is used to drive the blanket and it is produced frommore » the interaction of a 1-GeV proton beam with a lead-bismuth eutectic (LBE) target. The blanket has a liquid mobile fuel using LBE as the fuel carrier. The fuel materials are dissolved, mixed, or suspended in the liquid fuel carrier. Monte Carlo analyses were performed to determine the overall parameters of the concept. Steady-state Monte Carlo simulations were performed for three similar fission blankets. Except for, the loaded amount of actinide materials in the LBE is either 5, 7, or 10% of the total volume of the blanket, respectively. The neutron multiplication factors of the three blankets are ~0.98 and the initial MAs blanket inventories are ~10 tons. In addition, Monte Carlo burnup simulations using the MCB5 code were performed to analyze the performance of the three conceptual ADS systems. During operation, fresh fuel was fed into the fission blanket to adjust its reactivity and to control the system power. The burnup analysis shows that the three ADS concepts consume about 1.2 tons of actinides per full power year and produce 3 GW thermal power, with a proton beam power of 25 MW. For the blankets with 5, 7, or 10% actinide fuel particles loaded in the LBE, assuming that the ADS systems can be operated for 35 full-power years, the total MA materials consumed in the three ADS systems are about 30.6, 35.3, and 37.2 tons, respectively. Thus, the corresponding numbers of ADS systems to utilize the 131 tons of MA materials of the SNF inventory are 4.3, 3.7, or 3.5, respectively. ADS concepts with tube bundles inserted in the fission blanket were analyzed to overcome the disadvantages of the homogeneous blanket concept. The liquid lead is used as the target material, the mobile fuel carrier, and the primary coolant to avoid the polonium production from bismuth. Reactor physics and thermal-hydraulic analyses were coupled to determine the parameters of the heterogeneous fission blanket. The engineering requirements for a satisfactory operation performance of the HT-9 ferritic steel structure material have been realized. Two heterogeneous concepts of the subcritical fission blanket with the liquid lead mobile fuel inside or outside the tube bundles were considered. The heterogeneous configuration with the mobile fuel inside the tubes showed better performance than the configuration with mobile fuel outside the bundle tubes. The Monte Carlo burnup codes, MCB5 and SERPENT were both used to simulate the fuel burnup in the ADS concepts with the mobile fuels inside the tubes. The burnup analyses were carried out for 35 full power years. The results show that 5 ADS systems can dispose of the total United States inventory of the spent nuclear fuel.« less

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

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

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

Automated closure system for nuclear reactor fuel assemblies

DOEpatents

Christiansen, David W.; Brown, William F.

1985-01-01

A welder for automated closure of fuel pins by a pulsed magnetic process in which the open end of a length of cladding is positioned within a complementary tube surrounded by a pulsed magnetic welder. Seals are provided at each end of the tube, which can be evacuated or can receive tag gas for direct introduction to the cladding interior. Loading of magnetic rings and end caps is accomplished automatically in conjunction with the welding steps carried out within the tube.

Gas tungsten arc welder with electrode grinder

DOEpatents

Christiansen, David W.; Brown, William F.

1984-01-01

A welder for automated closure of fuel pins by a gas tungsten arc process in which a rotating length of cladding is positioned adjacent a welding electrode in a sealed enclosure. An independently movable axial grinder is provided in the enclosure for refurbishing the used electrode between welds.

The Pitt Innovation Challenge (PInCh): Driving Innovation in Translational Research Through an Incentive-Based, Problem-Focused Competition

PubMed Central

Fitzpatrick, Nicole Edgar; Maier, John; Yasko, Laurel; Mathias, David; Qua, Kacy; Wagner, Erika; Miller, Elizabeth; Reis, Steven E.

2017-01-01

Problem Translational research aims to move scientific discoveries across the biomedical spectrum from the laboratory to humans, and to ultimately transform clinical practice and public health policies. Despite efforts to accelerate translational research through national initiatives, several major hurdles remain. Approach The authors created the Pitt Innovation Challenge (PInCh) as an incentive-based, problem-focused approach to solving identified clinical or public health problems at the University of Pittsburgh Clinical and Translational Science Institute in spring 2014. With input from a broad range of stakeholders, PInCh leadership arrived at the challenge question: How do we empower individuals to take control of their own health outcomes? The authors developed the PInCh’s three-round proposal submission and review process as well as an online contest management tool to support the process. Outcomes Ninety-two teams submitted videos proposals in round one. Proposals included mobile applications (29, 32%), other information technology (19, 21%), and community program (22, 24%) solutions. Ten teams advanced to the final round, where three were awarded $100,000 to implement their solution over twelve months. In a six-month follow-up survey, 6/11 (55%) team leaders stated the PInCh helped to facilitate connections outside their normal sphere of collaborators. Next Steps Additional educational training sessions related to problem-focused research will be developed. The PInCh will be expanded to engage investment and industry communities to facilitate the translation of solutions to clinical practice via commercialization pathways. External organizations and other universities will be engaged to use the PInCh as a mechanism to fuel innovation in their spaces. PMID:27508341

The Japanese utilities` expectations for subchannel analysis

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

Toba, Akio; Omoto, Akira

1995-12-01

Boiling water reactor (BWR) utilities in Japan began to consider the development of a mechanistic model to describe the critical heat transfer conditions in the BWR fuel subchannel. Such a mechanistic model will not only decrease the necessity of tests, but will also help by removing some overly conservative safety margins in thermal hydraulics. With the use of a postdryout heat transfer correlation, new acceptance criteria may be applicable to evaluate the fuel integrity. Mechanistic subchannel analysis models will certainly back up this approach. This model will also be applicable to the analysis of large-size fuel bundles and examination ofmore » corrosion behavior.« less

Source Term Experiments Project (STEP): Aerosol characterization system

NASA Astrophysics Data System (ADS)

Schlenger, B. J.; Dunn, P. F.

A series of four experiments is being conducted at Argonne National Laboratory's TREAT Reactor. They were designed to provide some of the necessary data regarding magnitude and release rates of fission products from degraded fuel pins, physical and chemical characteristics of released fission products, and aerosol formation and transport phenomena. These are in pile experiments, whereby the test fuel is heated by neutron induced fission and subsequent clad oxidation in steam environments that simulate as closely as practical predicted reactor accident conditions. The test sequences cover a range of pressure and fuel heatup rate, and include the effect of Aq/In/Cd control rod material.

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

DOEpatents

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

1998-04-21

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

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

DOEpatents

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

1998-01-01

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

Enhancing BWR proliferation resistance fuel with minor actinides

NASA Astrophysics Data System (ADS)

Chang, Gray S.

2009-03-01

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.

USING WASTE TO CLEAN UP THE ENVIRONMENT: CELLULOSIC ETHANOL, THE FUTURE OF FUELS

EPA Science Inventory

In the process of converting municipal solid waste (MSW) into ethanol we optimized the first two major steps of pretreatment and enzymatic hydrolysis stages to enhance the sugar yield and to reduce the cost. For the pretreatment process, we tested different parameters of react...

Manifold, bus support and coupling arrangement for solid oxide fuel cells

DOEpatents

Parry, G.W.

1988-04-21

Individual, tubular solid oxide fuel cells (SOFCs) are assembled into bundles called a module within a housing, with a plurality of modules arranged end-to-end in a linear, stacked configuration called a string. A common set of piping comprised of a suitable high temperature resistant material (1) provides fuel and air to each module housing, (2) serves as electrically conducting buses, and (3) provides structural support for a string of SOFC modules. Ceramic collars are used to connect fuel and air inlet piping to each of the electrodes in an SOFC module and provide (1) electrical insulation for the current carrying bus bars and gas manifolds, (2) damping for the fuel and air inlet piping, and (3) proper spacing between the fuel and air inlet piping to prevent contact between these tubes and possible damage to the SOFC. 11 figs.

Final report of fuel dynamics Test E7

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

Doerner, R.C.; Murphy, W.F.; Stanford, G.S.

1977-04-01

Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release (there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow); (3) the predominant postfailure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4)more » less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets.« less

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

Measurement of liquid film flow on nuclear rod bundle in micro-scale by using very high speed camera system

NASA Astrophysics Data System (ADS)

Pham, Son; Kawara, Zensaku; Yokomine, Takehiko; Kunugi, Tomoaki

2012-11-01

Playing important roles in the mass and heat transfer as well as the safety of boiling water reactor, the liquid film flow on nuclear fuel rods has been studied by different measurement techniques such as ultrasonic transmission, conductivity probe, etc. Obtained experimental data of this annular two-phase flow, however, are still not enough to construct the physical model for critical heat flux analysis especially at the micro-scale. Remain problems are mainly caused by complicated geometry of fuel rod bundles, high velocity and very unstable interface behavior of liquid and gas flow. To get over these difficulties, a new approach using a very high speed digital camera system has been introduced in this work. The test section simulating a 3×3 rectangular rod bundle was made of acrylic to allow a full optical observation of the camera. Image data were taken through Cassegrain optical system to maintain the spatiotemporal resolution up to 7 μm and 20 μs. The results included not only the real-time visual information of flow patterns, but also the quantitative data such as liquid film thickness, the droplets' size and speed distributions, and the tilt angle of wavy surfaces. These databases could contribute to the development of a new model for the annular two-phase flow. Partly supported by the Global Center of Excellence (G-COE) program (J-051) of MEXT, Japan.

Heat transfer enhancement with mixing vane spacers using the field synergy principle

NASA Astrophysics Data System (ADS)

Yang, Lixin; Zhou, Mengjun; Tian, Zihao

2017-01-01

The single-phase heat transfer characteristics in a PWR fuel assembly are important. Many investigations attempt to obtain the heat transfer characteristics by studying the flow features in a 5 × 5 rod bundle with a spacer grid. The field synergy principle is used to discuss the mechanism of heat transfer enhancement using mixing vanes according to computational fluid dynamics results, including a spacer grid without mixing vanes, one with a split mixing vane, and one with a separate mixing vane. The results show that the field synergy principle is feasible to explain the mechanism of heat transfer enhancement in a fuel assembly. The enhancement in subchannels is more effective than on the rod's surface. If the pressure loss is ignored, the performance of the split mixing vane is superior to the separate mixing vane based on the enhanced heat transfer. Increasing the blending angle of the split mixing vane improves heat transfer enhancement, the maximum of which is 7.1%. Increasing the blending angle of the separate mixing vane did not significantly enhance heat transfer in the rod bundle, and even prevented heat transfer at a blending angle of 50°. This finding testifies to the feasibility of predicting heat transfer in a rod bundle with a spacer grid by field synergy, and upon comparison with analyzed flow features only, the field synergy method may provide more accurate guidance for optimizing the use of mixing vanes.

Supercritical fluid attachment of palladium nanoparticles on aligned carbon nanotubes.

PubMed

Ye, Xiang-Rong; Lin, Yuehe; Wai, Chien M; Talbot, Jan B; Jin, Sungho

2005-06-01

Nanocomposite materials consisting of Pd nanoparticles deposited on aligned multi-walled carbon nanotubes have been fabricated through hydrogen reduction of palladium-beta-diketone precursor in supercritical carbon dioxide. The supercritical fluid processing allowed deposition of high-density Pd nanoparticles (approximately 5-10 nm) on both as-grown (unfunctionalized) and functionalized (using HNO3 oxidation) nanotubes. However, the wet processing for functionalization results in pre-agglomerated, bundle-shaped nanotubes, thus significantly reducing the effective surface area for Pd particle deposition, although the bundling provides more secure, lock-in-place positioning of nanotubes and Pd catalyst particles. The nanotube bundling is substantially mitigated by Pd nanoparticle deposition on the unfunctionalized and geometrically separated nanotubes, which provides much higher catalyst surface area. Such nanocomposite materials utilizing geometrically secured and aligned nanotubes can be useful for providing much enhanced catalytic activities to chemical and electrochemical reactions (e.g., fuel cell reactions), and eliminate the need for tedious catalyst recovery process after the reaction is completed.

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

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

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

2016-10-28

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

75 FR 43097 - Airworthiness Directives; The Boeing Company Model 757 Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-23

... must be sealed for lightning strike protection. Relevant Service Information AD 2008-23-19 referred to... additional fasteners in the main fuel tanks must be sealed for lightning strike protection. The Federal... bundles inside the left and right equipment cooling system bays, on the left and right rear spars, and on...

Catalytic membranes for fuel cells

DOEpatents

Liu, Di-Jia [Naperville, IL; Yang, Junbing [Bolingbrook, IL; Wang, Xiaoping [Naperville, IL

2011-04-19

A fuel cell of the present invention comprises a cathode and an anode, one or both of the anode and the cathode including a catalyst comprising a bundle of longitudinally aligned graphitic carbon nanotubes including a catalytically active transition metal incorporated longitudinally and atomically distributed throughout the graphitic carbon walls of said nanotubes. The nanotubes also include nitrogen atoms and/or ions chemically bonded to the graphitic carbon and to the transition metal. Preferably, the transition metal comprises at least one metal selected from the group consisting of Fe, Co, Ni, Mn, and Cr.

Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution

DOEpatents

Tomczuk, Zygmunt; Miller, William E.; Wolson, Raymond D.; Gay, Eddie C.

1991-01-01

An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

Flexible fuel cell gas manifold system

DOEpatents

Cramer, Michael; Shah, Jagdish; Hayes, Richard P.; Kelley, Dana A.

2005-05-03

A fuel cell stack manifold system in which a flexible manifold body includes a pan having a central area, sidewall extending outward from the periphery of the central area, and at least one compound fold comprising a central area fold connecting adjacent portions of the central area and extending between opposite sides of the central area, and a sidewall fold connecting adjacent portions of the sidewall. The manifold system further includes a rail assembly for attachment to the manifold body and adapted to receive pins by which dielectric insulators are joined to the manifold assembly.

Diffusion Bonding Technology of Tungsten and SiC/SiC Composites for Nuclear Applications

NASA Astrophysics Data System (ADS)

Kishimoto, Hirotatsu; Shibayama, Tamaki; Abe, Takahiro; Shimoda, Kazuya; Kawamura, Satoshi; Kohyama, Akira

2011-10-01

Silicon carbide (SiC) is a candidate for the structural material in the next generation nuclear plants. Use of SiC/SiC composites is expected to increase the operation temperature of system over 1000 °C. For the high temperature system, refractory metals are planned to be used for several components. Tungsten is a candidate of armor on the divertor component in fusion, and is planned to be used for an upper-end plug of SiC/SiC fuel pin in a Gas cooled Fast Reactor (GFR). Joining technique of the SiC/SiC composites and tungsten is an important issue for nuclear systems in future. Nano-Infiltration and Transient Eutectoid (NITE) method is able to provide dense stable and high strength SiC/SiC composites having high resistance against pressure at elevated temperature, a diffusion bonding technique is usable to join the materials. Present research produces a NITE-SiC/SiC composite and tungsten as the similar dimension as a projected cladding tube of fuel pin for GFR using diffusion bonding, and investigated microstructure and mechanical properties.

Methodology for the study of the boiling crisis in a nuclear fuel bundle

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

Crecy, F. de; Juhel, D.

1995-09-01

The boiling crisis is one of the phenoumena limiting the available power from a nuclear power plant. It has been widely studied for decades, and numerous data, models, correlations or tables are now available in the literature. If we now try to obtain a general view of previous work in this field, we may note that there are several ways of tackling the subject. The mechanistic models try to model the two-phase flow topology and the interaction between different sublayers, and must be validated by comparison with basic experiments, such as DEBORA, where we try to obtain some detailed informationsmore » on the two-phase flow pattern in a pure and simple geometry. This allows us to obtain better knowledge of the so-called {open_quotes}intrinsic effect{close_quotes}. These models are not yet acceptable for nuclear use. As the geometry of the rod bundles and grids has a tremendous importance for the Critical Heat Flux (CHF), it is mandatory to have more precise results for a given fuel rod bundle in a restricted range of parameters: this leads to the empirical approach, using empirical CHF predictors (tables, correlations, splines, etc...). One of the key points of such a method is the obtaining local thermohydraulic values, that is to say the evaluation of the so-called {open_quotes}mixing effect{close_quotes}. This is done by a subchannel analysis code or equivalent, which can be qualified on two kinds of experiments: overall flow measurements in a subchannel, such as HYDROMEL in single-phase flow or GRAZIELLA in two-phase flow, or detailed measurements inside a subchannel, such as AGATE. Nevertheless, the final qualification of a specific nuclear fuel, i.e. the synthesis of these mechanistic and empirical approaches, intrinsic and mixing effects, etc..., must be achieved on a global test such as OMEGA. This is the strategy used in France by CEA and its partners FRAMATOME and EdF.« less

Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

DOE PAGES

Bakosi, J.; Christon, M. A.; Lowrie, R. B.; ...

2013-07-12

The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3 × 3 and 5 × 5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carriedmore » out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the singlephase incompressible Navier-Stokes equations. The simulations explicitly resolve the large scale motions of the turbulent flow field using first principles and rely on a monotonicity-preserving numerical technique to represent the unresolved scales. Each series of simulations for the 3 × 3 and 5 × 5 rod-bundle geometries is an analysis of the flow field statistics combined with a mesh-refinement study and validation with available experimental data. Our primary focus is the time history and statistics of the forces loading the fuel rods. These hydrodynamic forces are believed to be the key player resulting in rod vibration and GTRF wear, one of the leading causes for leaking nuclear fuel which costs power utilities millions of dollars in preventive measures. As a result, we demonstrate that implicit large-eddy simulation of rod-bundle flows is a viable way to calculate the excitation forces for the GTRF problem.« less

Irradiation subassembly

DOEpatents

Seim, O.S.; Filewicz, E.C.; Hutter, E.

1973-10-23

An irradiation subassembly for use in a nuclear reactor is described which includes a bundle of slender elongated irradiation -capsules or fuel elements enclosed by a coolant tube and having yieldable retaining liner between the irradiation capsules and the coolant tube. For a hexagonal bundle surrounded by a hexagonal tube the yieldable retaining liner may consist either of six segments corresponding to the six sides of the tube or three angular segments each corresponding in two adjacent sides of the tube. The sides of adjacent segments abut and are so cut that metal-tometal contact is retained when the volume enclosed by the retaining liner is varied and Springs are provided for urging the segments toward the center of the tube to hold the capsules in a closely packed configuration. (Official Gazette)

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

PubMed

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

2014-05-01

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

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)

Method for monitoring irradiated fuel using Cerenkov radiation

DOEpatents

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

1980-05-21

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

Scoping studies of vapor behavior during a severe accident in a metal-fueling reactor

NASA Astrophysics Data System (ADS)

Spencer, B. W.; Marchaterre, J. F.

1985-04-01

The consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel were examined. The principal gas and vapor species released are shown to be Xe, Cs, and bond sodium contained within the fuel porosity. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core. If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the ability of vapor expansion to perform appreciable work on the system and the ability of an expanding vapor bubble to transport fuel and fission produce species to the cover gas region where they may be released to the containment are largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool.

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

NASA Astrophysics Data System (ADS)

Najeeb, Umair

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

Evaluation of CASL boiling model for DNB performance in full scale 5x5 fuel bundle with spacer grids

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

Kim, Seung Jun

As one of main tasks for FY17 CASL-THM activity, Evaluation study on applicability of the CASL baseline boiling model for 5x5 DNB application is conducted and the predictive capability of the DNB analysis is reported here. While the baseline CASL-boiling model (GEN- 1A) approach has been successfully implemented and validated with a single pipe application in the previous year’s task, the extended DNB validation for realistic sub-channels with detailed spacer grid configurations are tasked in FY17. The focus area of the current study is to demonstrate the robustness and feasibility of the CASL baseline boiling model for DNB performance inmore » a full 5x5 fuel bundle application. A quantitative evaluation of the DNB predictive capability is performed by comparing with corresponding experimental measurements (i.e. reference for the model validation). The reference data are provided from the Westinghouse Electricity Company (WEC). Two different grid configurations tested here include Non-Mixing Vane Grid (NMVG), and Mixing Vane Grid (MVG). Thorough validation studies with two sub-channel configurations are performed at a wide range of realistic PWR operational conditions.« less

Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel

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

Melissa C. Teague; Brian P. Gorman; Steven L. Hayes

2013-10-01

High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column weremore » observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.« less

Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

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

Santamarina, A.; Bernard, D.; Blaise, P.

2013-07-01

This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency ofmore » Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)« less

Development of Innovative Accident Tolerant High Thermal Conductivity UO 2-Diamond Composite Fuel Pellets

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

Tulenko, James; Subhash, Ghatu

2016-01-01

The University of Florida (UF) evaluated a composite fuel consisting of UO 2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO 2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO 2 – SiC and UO 2 – Carbon Nanotube fuel pins. UF ismore » proving with the current research results that the addition of diamond micro particles to UO 2 may greatly enhanced the thermal conductivity of the UO 2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.« less

Computed tomography of radioactive objects and materials

NASA Astrophysics Data System (ADS)

Sawicka, B. D.; Murphy, R. V.; Tosello, G.; Reynolds, P. W.; Romaniszyn, T.

1990-12-01

Computed tomography (CT) has been performed on a number of radioactive objects and materials. Several unique technical problems are associated with CT of radioactive specimens. These include general safety considerations, techniques to reduce background-radiation effects on CT images and selection criteria for the CT source to permit object penetration and to reveal accurate values of material density. In the present paper, three groups of experiments will be described, for objects with low, medium and high levels of radioactivity. CT studies on radioactive specimens will be presented. They include the following: (1) examination of individual ceramic reactor-fuel (uranium dioxide) pellets, (2) examination of fuel samples from the Three Mile Island reactor, (3) examination of a CANDU (CANada Deuterium Uraniun: registered trademark) nuclear-fuel bundle which underwent a simulated loss-of-coolant accident resulting in high-temperature damage and (4) examination of a PWR nuclear-reactor fuel assembly.

Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

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

Jeong, Hae-Yong; Ha, Kwi-Seok; Chang, Won-Pyo

The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal-cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the descriptionmore » of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.« less

Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility

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

Doerner, R.C.; Bauer, T.H.; Morman, J.A.

Prototypic oxide fuel was subjected to simulated, fast reactor severe accident conditions in a series of in-pile tests in the Transient Reactor Test Facility reactor. Seven experiments were performed on fresh and previously irradiated oxide fuel pins under transient overpower and transient undercooled. overpower accident conditions. For each of the tests, fuel motions were observed by the hodoscope. Hodoscope data are correlated with coolant flow, pressure, and temperature data recorded by the loop instrumentation. Data were analyzed from the onset of initial failure to a final mass distribution at the end of the test. In this paper results of thesemore » analyses are compared to pre- and posttest accident calculations and to posttest metallographic accident calculations and to posttest metallographic examinations and computed tomographic reconstructions from neutron radiographs.« less

Relative sliding durability of candidate high temperature fiber seal materials

NASA Technical Reports Server (NTRS)

Dellacorte, Christopher; Steinetz, Bruce M.

1992-01-01

The relative sliding durability behavior of six candidate ceramic fibers for high temperature sliding seal applications is reviewed and compared. Pin on disk tests were used to evaluate potential seal materials by sliding a tow or bundle of the candidate ceramic fiber against a superalloy test disk. Tests were conducted in air under a 2.65 N load, at a sliding velocity of 0.025 m/sec and at temperatures from 25 to 900 C. Friction was measured during the tests and fiber wear, indicated by the extent of fibers broken in the tow or bundle, was measured at the end of each test. For most of the fibers, friction and wear increase with test temperature. The relative fiber durability ranking correlates with tensile strength, indicating that tensile data, which is more readily available than sliding durability data, may be useful in predicting fiber wear behavior under various conditions. A dimensional analysis of the wear data shows that the fiber durability is related to a dimensionless durability ratio which represents the ratio of the fiber strength to the fiber stresses imposed by sliding. The analysis is applicable to fibers with similar diameters and elastic moduli. Based upon the results of the research program, three fiber candidates are recommended for further study as potential seal materials. They are a silicon based complex carbide-oxide fiber, an alumina-boria-silica and an aluminosilicate fiber.

Iso-geometric analysis for neutron diffusion problems

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

Hall, S. K.; Eaton, M. D.; Williams, M. M. R.

Iso-geometric analysis can be viewed as a generalisation of the finite element method. It permits the exact representation of a wider range of geometries including conic sections. This is possible due to the use of concepts employed in computer-aided design. The underlying mathematical representations from computer-aided design are used to capture both the geometry and approximate the solution. In this paper the neutron diffusion equation is solved using iso-geometric analysis. The practical advantages are highlighted by looking at the problem of a circular fuel pin in a square moderator. For this problem the finite element method requires the geometry tomore » be approximated. This leads to errors in the shape and size of the interface between the fuel and the moderator. In contrast to this iso-geometric analysis allows the interface to be represented exactly. It is found that, due to a cancellation of errors, the finite element method converges more quickly than iso-geometric analysis for this problem. A fuel pin in a vacuum was then considered as this problem is highly sensitive to the leakage across the interface. In this case iso-geometric analysis greatly outperforms the finite element method. Due to the improvement in the representation of the geometry iso-geometric analysis can outperform traditional finite element methods. It is proposed that the use of iso-geometric analysis on neutron transport problems will allow deterministic solutions to be obtained for exact geometries. Something that is only currently possible with Monte Carlo techniques. (authors)« less

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

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

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

2015-01-01

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

Emergency cooling analysis for the loss of coolant malfunction

NASA Technical Reports Server (NTRS)

Peoples, J. A.

1972-01-01

This report examines the dynamic response of a conceptual space power fast-spectrum lithium cooled reactor to the loss of coolant malfunction and several emergency cooling concepts. The results show that, following the loss of primary coolant, the peak temperatures of the center most 73 fuel elements can range from 2556 K to the region of the fuel melting point of 3122 K within 3600 seconds after the start of the accident. Two types of emergency aftercooling concepts were examined: (1) full core open loop cooling and (2) partial core closed loop cooling. The full core open loop concept is a one pass method of supplying lithium to the 247 fuel pins. This method can maintain fuel temperature below the 1611 K transient damage limit but requires a sizable 22,680-kilogram auxiliary lithium supply. The second concept utilizes a redundant internal closed loop to supply lithium to only the central area of each hexagonal fuel array. By using this method and supplying lithium to only the triflute region, fuel temperatures can be held well below the transient damage limit.

Lateral Structure Formation in Polyelectrolyte Brushes Induced by Multivalent Ions

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

Brettmann, Blair; Pincus, Philip; Tirrell, Matthew

2017-01-13

We provide a theoretical model for the collapse of polyelectrolyte brushes in the presence of multivalent ions, focusing on the formation of lateral inhomogeneties in the collapsed state. Polyelectrolyte brushes are important in a variety of applications, including stabilizing colloidal particles and lubricating surfaces. Many uses rely on the extension of the densely grafted polymer chains from the surface in the extended brush morphology. In the presence Extended Brush of multivalent ions, brushes are significantly shorter than in monovalent ionic solutions, which greatly affects their properties. We base our theoretical analysis on an analogous collapse of polyelectrolyte brushes in amore » poor solvent, providing an energy balance representation for pinned micelles and cylindrical bundles. The equilibrium brush heights predicted for these structures are of a similar magnitude to those measured experimentally. The formation of lateral structures can open new avenues for stimuli-responsive applications that rely on nanoscale pattern formation on surfaces.« less

Funnel for fuel pin loading system

DOEpatents

Christiansen, D.W.; Steffen, J.M.; Brown, W.F.

1984-01-01

An enlarged funnel is described which is releasably mounted at the open end of a length of cladding by an encircling length of shrink tubing which securely engages outer surfaces of both the funnel and cladding. The shrink tubing overlaps an annular shoulder against which pulling force can be exerted to remove the tubing from the cladding. The shoulder can be provided on a separate collar or ring, or on the funnel itself.

Funnel for fuel pin loading system

DOEpatents

Christiansen, David W.; Steffen, Jim M.; Brown, William F.

1985-01-01

An enlarged funnel is releasably mounted at the open end of a length of cladding by an encircling length of shrink tubing which securely engages outer surfaces of both the funnel and cladding. The shrink tubing overlaps an annular shoulder against which pulling force can be exerted to remove the tubing from the cladding. The shoulder can be provided on a separate collar or ring, or on the funnel itself.

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to make scoping calculations for mission analysis.

The Effect of Stress and Hot Corrosion on Nickel-Base Superalloys

DTIC Science & Technology

1985-03-01

in a degradation of material properties and reduced component life. Allen and Whitlow(6). stated that superalloys in combustion turbine environments...pins are tested in combustion gas streams at elevated temperatures. A hot corrosion environment is usually simulated by burning a sulfur-containing fuel...corrosion attack frequently observed on combustion turbine blades retrieved from service. Figure 1 shows the effect of salt thickness on hot corrosion

Calculation of Internal Pressures in the Fuel Tube of a Nuclear Reactor

NASA Technical Reports Server (NTRS)

Rosenbaum, B. M.; Allen, G.

1952-01-01

General procedures for computing internal pressures in fuel tubes of nuclear reactors are described and the effects on the pressure of varying neutron flux, fissioning material, and operating temperatures are discussed. A general proof is given that during pile operation each fission product is monotonically increasing and therefore a maximum amount of all elements is present at the time of shit down. The post-shutdown build-up of elements that are held in check during pile operation because of their inordinately high capture cross sections is calculated quantitatively. An account of chemical interactions between the many fission-product elements and the resulting effect on the total pressure completes the discussion. The general methods are illustrated by calculations applied to a system consisting of 90 percent enriched U235 in the form of UO2 packed into a hollow metal cylinder or "pin", operating at a flux of 8 x 10(exp 14) at 2000 F. Calculations of the pressure inside a pin are made with and without a sodium metal heat-transfer additive. The bulk of the pressure is shown to depend on the four elements, xenon, krypton, rubidium, and cesium; the amount of free oxygen, however, was also significant. For a shutdown time of 10(exp 6) seconds, the pressure was about 100 atmospheres.

Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor.

PubMed

Košťál, Michal; Švadlenková, Marie; Koleška, Michal; Rypar, Vojtěch; Milčák, Ján

2015-11-01

Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins. Copyright © 2015 Elsevier Ltd. All rights reserved.

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

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

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

A Nuclear Reactor and Chemical Processing Design for Production of Molybdenum-99 with Crystalline Uranyl Nitrate Hexahydrate Fuel

NASA Astrophysics Data System (ADS)

Stange, Gary Michael

Medical radioisotopes are used in tens of millions of procedures every year to detect and image a wide variety of maladies and conditions in the human body. The most widely-used diagnostic radioisotope is technetium-99m, a metastable isomer of technetium-99 that is generated by the radioactive decay of molybdenum-99. For a number of reasons, the supply of molybdenum-99 has become unreliable and the techniques used to produce it have become unattractive. This has spurred the investigation of new technologies that avoid the use of highly enriched uranium to produce molybdenum-99 in the United States, where approximately half of the demand originates. The first goal of this research is to develop a critical nuclear reactor design powered by solid, discrete pins of low enriched uranium. Analyses of single-pin heat transfer and whole-core neutronics are performed to determine the required specifications. Molybdenum-99 is produced directly in the fuel of this reactor and then extracted through a series of chemical processing steps. After this extraction, the fuel is left in an aqueous state. The second goal of this research is to describe a process by which the uranium may be recovered from this spent fuel solution and reconstituted into the original fuel form. Fuel recovery is achieved through a crystallization step that generates solid uranyl nitrate hexahydrate while leaving the majority of fission products and transuranic isotopes in solution. This report provides background information on molybdenum-99 production and crystallization chemistry. The previously unknown thermal conductivity of the fuel material is measured. Following this is a description of the modeling and calculations used to develop a reactor concept. The operational characteristics of the reactor core model are analyzed and reported. Uranyl nitrate crystallization experiments have also been conducted, and the results of this work are presented here. Finally, a process flow scheme for uranium recovery is examined, in part qualitatively and in part quantitatively, based upon the preceding data garnered through literature review, modeling, and experimentation. The sum of this research is meant to allow for a complete understanding of the process flow, from the beginning of one production cycle to the beginning of another.

Heat transfer in laminar flow along circular rods in infinite square arrays

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

Kim, J.H.; Li, W.H.

1988-02-01

The need to understand heat transfer characteristics over rods or tube bundles often arises in the design of compact heat exchangers and safety analysis of nuclear reactors. In particular, the fuel bundles of typical light water nuclear reactors are composed of a large number of circular rods arranged in square array pattern. The purpose of the present study is to analyze heat transfer characteristics of flow in such a multirod geometric configuration. The analysis given here will follow as closely as possible the method of Sparrow et al. who analyzed a similar problem for circular cylinders arranged in an equilateralmore » triangular array. The following major assumptions are made in the present analysis: (1) Flow is fully developed laminar flow paralleled to the axis of rods. (2) The axial profile of the surface heat flux to the fluid is uniform.(3) Thermodynamic properties are assumed constant.« less

Nuclear Resonance Fluorescence for Materials Assay

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

Quiter, Brian; Ludewigt, Bernhard; Mozin, Vladimir

This paper discusses the use of nuclear resonance fluorescence (NRF) techniques for the isotopic and quantitative assaying of radioactive material. Potential applications include age-dating of an unknown radioactive source, pre- and post-detonation nuclear forensics, and safeguards for nuclear fuel cycles Examples of age-dating a strong radioactive source and assaying a spent fuel pin are discussed. The modeling work has ben performed with the Monte Carlo radiation transport computer code MCNPX, and the capability to simulate NRF has bee added to the code. Discussed are the limitations in MCNPX's photon transport physics for accurately describing photon scattering processes that are importantmore » contributions to the background and impact the applicability of the NRF assay technique.« less

Multidimensional effects in the thermal response of fuel rod simulators. [PWR

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

Dabbs, R.D.; Ott, L.J.

1980-01-01

One of the primary objectives of the Oak Ridge National Laboratory Pressurized-Water Reactor Blowdown Heat Transfer Separate-Effects Program is the determination of the transient surface temperature and surface heat flux of fuel pin simulators (FPSs) from internal thermocouple signals obtained during a loss-of-coolant experiment (LOCE) in the Thermal-Hydraulics Test Facility. This analysis requires the solution of the classical inverse heat conduction problem. The assumptions that allow the governing differential equation to be reduced to one dimension can introduce significant errors in the computed surface heat flux and surface temperature. The degree to which these computed variables are perturbed is addressedmore » and quantified.« less

CRITICAL EXPERIMENT WITH BORAX-V. Internal Superheater

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

Plumlee, K.E.; Baird, Q.L.; Stanford, G.S.

1963-11-01

A critical experiment was performed with 12 BORAX-V superheater subassemblies in a central voidable region plus 1228 to 1525 UO/sub 2/ fuel pins (3 wt% enriched) in a peripheral region. Removing water (28% of superheater volume) at room temperature decreased reactivity by 2.2%. The midplane (two- dimensional) peak-to-average power distribution in the voided superheater was approximately 1.24, mostly attributable to flux depressions within insulated fuel boxes. Cadmium ratios are also reported. The experiment was initiated to supplement computational information which might have affected plans for loading the superheater zone into the BORAX-V reactor. No changes were indicated by the experiment.more » (auth)« less

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

A Full-Core Resonance Self-Shielding Method Using a Continuous-Energy Quasi–One-Dimensional Slowing-Down Solution that Accounts for Temperature-Dependent Fuel Subregions and Resonance Interference

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

Liu, Yuxuan; Martin, William; Williams, Mark

In this paper, a correction-based resonance self-shielding method is developed that allows annular subdivision of the fuel rod. The method performs the conventional iteration of the embedded self-shielding method (ESSM) without subdivision of the fuel to capture the interpin shielding effect. The resultant self-shielded cross sections are modified by correction factors incorporating the intrapin effects of radial variation of the shielded cross section, radial temperature distribution, and resonance interference. A quasi–one-dimensional slowing-down equation is developed to calculate such correction factors. The method is implemented in the DeCART code and compared with the conventional ESSM and subgroup method with benchmark MCNPmore » results. The new method yields substantially improved results for both spatially dependent reaction rates and eigenvalues for typical pressurized water reactor pin cell cases with uniform and nonuniform fuel temperature profiles. Finally, the new method is also proved effective in treating assembly heterogeneity and complex material composition such as mixed oxide fuel, where resonance interference is much more intense.« less

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

DOE PAGES

Philip, Bobby; Berrill, Mark A.; Allu, Srikanth; ...

2015-01-26

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

Tribological evaluation of an Al2O3-SiO2 ceramic fiber candidate for high temperature sliding seals

NASA Technical Reports Server (NTRS)

Dellacorte, Christopher; Steinetz, Bruce

1992-01-01

A test program to determine the relative sliding durability of an alumina-silica candidate ceramic fiber for high temperature sliding seal applications as described. This work represents the first reporting of the sliding durability of this material system. Pin-on-disk tests were used to evaluate the potential seal material by sliding a tow or bundle of the candidate ceramic fiber against a superalloy test disk. Friction was measured during the tests and fiber wear, indicated by the extent of fibers broken in the tow or bundle, was measured at the end of each test. Test variables studied included ambient temperatures from 25 C to 900 C, loads from 1.3 to 21.2 Newtons, and sliding velocities from 0.025 to 0.25 m/sec. In addition, the effects of fiber diameter, elastic modulus, and a pretest fiber heat treatment on friction and wear were measured. In most cases, wear increased with temperature. Friction ranged from about 0.36 at 500 C and low velocity (0.025 m/s) to over 1.1 at 900 C and high velocity (0.25 m/s). The pretest fiber heat treatment, which caused significant durability reductions for alumina-boria-silica ceramic fibers tested previously, had little effect on the alumina-silica fibers tested here. These results indicate that the alumina-silica (Al2O3-SiO2) fiber is a good candidate material system for high temperature sliding seal applications.

Tribological evaluation of an Al2O3-SiO2 ceramic fiber candidate for high temperature sliding seals

NASA Technical Reports Server (NTRS)

Dellacorte, Christopher; Steinetz, Bruce

1994-01-01

A test program to determine the relative sliding durability of an alumina-silica candidate ceramic fiber for high temperature sliding seal applications is described. Pin-on-disk tests were used to evaluate the potential seal material by sliding a tow or bundle of the candidate ceramic fiber against a superalloy test disk. Friction was measured during the tests and fiber wear, indicated by the extent of fibers broken in the tow or bundle, was measured at the end of each test. Test variables studied included ambient temperatures from 25 to 900 C, loads from 1.3 to 21.2 N, and sliding velocities from 0.025 to 0.25 m/sec. In addition, the effects of fiber diameter and elastic modulus on friction and wear were measured. Thin gold films deposited on the superalloy disk surface were evaluated in an effort to reduce friction and wear of the fibers. In most cases, wear increased with test temperature. Friction ranged from 0.36 at 500 C and low velocity (0.025 m/sec) to over 1.1 at 900 C and high velocity (0.25 m/sec). The gold films resulted in satisfactory lubrication of the fibers at 25 C. At elevated temperatures diffusion of substrate elements degraded the films. These results indicate that the alumina-silica (Al2O3-SiO2) fiber is a good candidate material system for high temperature sliding seal applications. More work is needed to reduce friction.

Data trove helps pin down the shape of the Milky Way

NASA Astrophysics Data System (ADS)

Clery, Daniel

2018-04-01

On 25 April, hundreds of astronomers around the world got their hands on one of the biggest data dumps in the history of astronomy: the exact positions, motions, brightnesses, and colors of 1.3 billion stars in and around the Milky Way, gathered during the first 2 years of operation by the European Space Agency's €750 million Gaia satellite, launched in 2013. The data—Gaia's second release—will fuel an explosion of studies about the structure of the Milky Way. The trove of stellar positions from Gaia will sharpen our picture of the galaxy's spiral structures, and the data on stellar motions will allow astronomers to wind the clock backward and see how the galaxy evolved over the past 13 billion years. Gaia's color and brightness information, meanwhile, will help astronomers classify stars by composition and pin down where different types are born.

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

Gleicher, Frederick; Ortensi, Javier; DeHart, Mark

Accurate calculation of desired quantities to predict fuel behavior requires the solution of interlinked equations representing different physics. Traditional fuels performance codes often rely on internal empirical models for the pin power density and a simplified boundary condition on the cladding edge. These simplifications are performed because of the difficulty of coupling applications or codes on differing domains and mapping the required data. To demonstrate an approach closer to first principles, the neutronics application Rattlesnake and the thermal hydraulics application RELAP-7 were coupled to the fuels performance application BISON under the master application MAMMOTH. A single fuel pin was modeledmore » based on the dimensions of a Westinghouse 17x17 fuel rod. The simulation consisted of a depletion period of 1343 days, roughly equal to three full operating cycles, followed by a station blackout (SBO) event. The fuel rod was depleted for 1343 days for a near constant total power loading of 65.81 kW. After 1343 days the fission power was reduced to zero (simulating a reactor shut-down). Decay heat calculations provided the time-varying energy source after this time. For this problem, Rattlesnake, BISON, and RELAP-7 are coupled under MAMMOTH in a split operator approach. Each system solves its physics on a separate mesh and, for RELAP-7 and BISON, on only a subset of the full problem domain. Rattlesnake solves the neutronics over the whole domain that includes the fuel, cladding, gaps, water, and top and bottom rod holders. Here BISON is applied to the fuel and cladding with a 2D axi-symmetric domain, and RELAP-7 is applied to the flow of the circular outer water channel with a set of 1D flow equations. The mesh on the Rattlesnake side can either be 3D (for low order transport) or 2D (for diffusion). BISON has a matching ring structure mesh for the fuel so both the power density and local burn up are copied accurately from Rattlesnake. At each depletion time step, Rattlesnake calculates a power density, fission density rate, burn-up distribution and fast flux based on the current water density and fuel temperature. These are then mapped to the BISON mesh for a fuels performance solve. BISON calculates the fuel temperature and cladding surface temperature based upon the current power density and bulk fluid temperature. RELAP-7 then calculates the fluid temperature, water density fraction and water phase velocity based upon the cladding surface temperature. The fuel temperature and the fluid density are then passed back to Rattlesnake for another neutronics calculation. Six Picard or fixed-point style iterations are preformed in this manner to obtain consistent tightly coupled and stable results. For this paper a set of results from the detailed calculation are provided for both during depletion and the SBO event. We demonstrate that a detailed calculation closer to first principles can be done under MAMMOTH between different applications on differing domains.« less

Development of a plume-in-grid model for industrial point and volume sources: application to power plant and refinery sources in the Paris region

NASA Astrophysics Data System (ADS)

Kim, Y.; Seigneur, C.; Duclaux, O.

2014-04-01

Plume-in-grid (PinG) models incorporating a host Eulerian model and a subgrid-scale model (usually a Gaussian plume or puff model) have been used for the simulations of stack emissions (e.g., fossil fuel-fired power plants and cement plants) for gaseous and particulate species such as nitrogen oxides (NOx), sulfur dioxide (SO2), particulate matter (PM) and mercury (Hg). Here, we describe the extension of a PinG model to study the impact of an oil refinery where volatile organic compound (VOC) emissions can be important. The model is based on a reactive PinG model for ozone (O3), which incorporates a three-dimensional (3-D) Eulerian model and a Gaussian puff model. The model is extended to treat PM, with treatments of aerosol chemistry, particle size distribution, and the formation of secondary aerosols, which are consistent in both the 3-D Eulerian host model and the Gaussian puff model. Furthermore, the PinG model is extended to include the treatment of volume sources to simulate fugitive VOC emissions. The new PinG model is evaluated over Greater Paris during July 2009. Model performance is satisfactory for O3, PM2.5 and most PM2.5 components. Two industrial sources, a coal-fired power plant and an oil refinery, are simulated with the PinG model. The characteristics of the sources (stack height and diameter, exhaust temperature and velocity) govern the surface concentrations of primary pollutants (NOx, SO2 and VOC). O3 concentrations are impacted differently near the power plant than near the refinery, because of the presence of VOC emissions at the latter. The formation of sulfate is influenced by both the dispersion of SO2 and the oxidant concentration; however, the former tends to dominate in the simulations presented here. The impact of PinG modeling on the formation of secondary organic aerosol (SOA) is small and results mostly from the effect of different oxidant concentrations on biogenic SOA formation. The investigation of the criteria for injecting plumes into the host model (fixed travel time and/or puff size) shows that a size-based criterion is recommended to treat the formation of secondary aerosols (sulfate, nitrate, and ammonium), in particular, farther downwind of the sources (beyond about 15 km). The impacts of PinG modeling are less significant in a simulation with a coarse grid size (10 km) than with a fine grid size (2 km), because the concentrations of the species emitted from the PinG sources are relatively less important compared to background concentrations when injected into the host model with a coarser grid size.

Development of a plume-in-grid model for industrial point and volume sources: application to power plant and refinery sources in the Paris region

NASA Astrophysics Data System (ADS)

Kim, Y.; Seigneur, C.; Duclaux, O.

2013-11-01

Plume-in-grid (PinG) models incorporating a host Eulerian model and a subgrid-scale model (usually a Gaussian plume or puff model) have been used for the simulations of stack emissions (e.g., fossil fuel-fired power plants and cement plants) for gaseous and particulate species such as nitrogen oxides (NOx), sulfur dioxide (SO2), particulate matter (PM) and mercury (Hg). Here, we describe the extension of a PinG model to study the impact of an oil refinery where volatile organic compound (VOC) emissions can be important. The model is based on a reactive PinG model for ozone (O3), which incorporates a three-dimensional (3-D) Eulerian model and a Gaussian puff model. The model is extended to treat PM, with treatments of aerosol chemistry, particle size distribution, and the formation of secondary aerosols, which are consistent in both the 3-D Eulerian host model and the Gaussian puff model. Furthermore, the PinG model is extended to include the treatment of volume sources to simulate fugitive VOC emissions. The new PinG model is evaluated over Greater Paris during July 2009. Model performance is satisfactory for O3, PM2.5 and most PM2.5 components. Two industrial sources, a coal-fired power plant and an oil refinery, are simulated with the PinG model. The characteristics of the sources (stack height and diameter, exhaust temperature and velocity) govern the surface concentrations of primary pollutants (NOx, SO2 and VOC). O3 concentrations are impacted differently near the power plant than near the refinery, because of the presence of VOC emissions at the latter. The formation of sulfate is influenced by both the dispersion of SO2 and the oxidant concentration; however, the former tends to dominate in the simulations presented here. The impact of PinG modeling on the formation of secondary organic aerosols (SOA) is small and results mostly from the effect of different oxidant concentrations on biogenic SOA formation. The investigation of the criteria for injecting plumes into the host model (fixed travel time and/or puff size) shows that a size-based criterion is recommended to treat the formation of secondary aerosols (sulfate, nitrate, and ammonium), in particular, farther downwind of the sources (from about 15 km). The impacts of the PinG modeling are less significant in a simulation with a coarse grid size (10 km) than with a fine grid size (2 km), because the concentrations of the species emitted from the PinG sources are relatively less important compared to background concentrations when injected into the host model.

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

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

Sin Kim; Goon-Cherl Park

1995-09-01

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

Development of Advanced Ods Ferritic Steels for Fast Reactor Fuel Cladding

NASA Astrophysics Data System (ADS)

Ukai, S.; Oono, N.; Ohtsuka, S.; Kaito, T.

Recent progress of the 9CrODS steel development is presented focusing on their microstructure control to improve sufficient high-temperature strength as well as cladding manufacturing capability. The martensitic 9CrODS steel is primarily candidate cladding materials for the Generation IV fast reactor fuel. They are the attractive composite-like materials consisting of the hard residual ferrite and soft tempered martensite, which are able to be easily controlled by α-γ phase transformation. The residual ferrite containing extremely nanosized oxide particles leads to significantly improved creep rupture strength in 9CrODS cladding. The creep strength stability at extended time of 60,000 h at 700 ºC is ascribed to the stable nanosized oxide particles. It was also reviewed that 9CrODS steel has well irradiation stability and fuel pin irradiation test was conducted up to 12 at% burnup and 51 dpa at the cladding temperature of 700ºC.

Manifold, bus support and coupling arrangement for solid oxide fuel cells

DOEpatents

Parry, Gareth W.

1989-01-01

Individual, tubular solid oxide fuel cells (SOFCs) are assembled into bundles called a module within a housing, with a plurality of modules arranged end-to-end in a linear, stacked configuration called a string. A common set of piping comprised of a suitable high temperture resistant material (1) provides fuel and air to each module housing, (2) serves as electrically conducting buses, and (3) provides structural support for a string of SOFC modules. The piping thus forms a manfold for directing fuel and air to each module in a string and makes electrical contact with the module's anode and cathode to conduct the DC power generated by the SOFC. The piping also provides structureal support for each individual module and maintains each string of modules as a structurally integral unit for ensuring high strength in a large 3-dimensional array of SOFC modules. Ceramic collars are used to connect fuel and air inlet piping to each of the electrodes in an SOFC module and provide (1) electrical insulation for the current carrying bus bars and gas manifolds, (2) damping for the fuel and air inlet piping, and (3) proper spacing between the fuel and air inlet piping to prevent contact between these tubes and possible damage to the SOFC.

Development of a new bench for puncturing of irradiated fuel rods in STAR hot laboratory

NASA Astrophysics Data System (ADS)

Petitprez, B.; Silvestre, P.; Valenza, P.; Boulore, A.; David, T.

2018-01-01

A new device for puncturing of irradiated fuel rods in commercial power plants has been designed by Fuel Research Department of CEA Cadarache in order to provide experimental data of high precision on fuel pins with various designs. It will replace the current set-up that has been used since 1998 in hot cell 2 of STAR facility with more than 200 rod puncturing experiments. Based on this consistent experimental feedback, the heavy-duty technique of rod perforation by clad punching has been preserved for the new bench. The method of double expansion of rod gases is also retained since it allows upgrading the confidence interval of volumetric results obtained from rod puncturing. Furthermore, many evolutions have been introduced in the new design in order to improve its reliability, to make the maintenance easier by remote handling and to reduce experimental uncertainties. Tightness components have been studied with Sealing Laboratory Maestral at Pierrelatte so as to make them able to work under mixed pressure conditions (from vacuum at 10-5 mbar up to pressure at 50 bars) and to lengthen their lifetime under permanent gamma irradiation in hot cell. Bench ergonomics has been optimized to make its operating by remote handling easier and to secure the critical phases of a puncturing experiment. A high pressure gas line equipped with high precision pressure sensors out of cell can be connected to the bench in cell for calibration purposes. Uncertainty analyses using Monte Carlo calculations have been performed in order to optimize capacity of the different volumes of the apparatus according to volumetric characteristics of the rod to be punctured. At last this device is composed of independent modules which allow puncturing fuel pins out of different geometries (PWR, BWR, VVER). After leak tests of the device and remote handling simulation in a mock-up cell, several punctures of calibrated specimens have been performed in 2016. The bench will be implemented soon in hot cell 2 of STAR facility for final qualification tests. PWR rod punctures are already planned for 2018.

Multiscale Methods for Nuclear Reactor Analysis

NASA Astrophysics Data System (ADS)

Collins, Benjamin S.

The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques, that use diffusion theory and pin power reconstruction (PPR). Two different multiscale methods were developed and analyzed; the post-refinement multiscale method and the embedded multiscale method. The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is "post-refinement" and thus has no impact on the global solution. The embedded multiscale method allows the local solver to change the global solution to provide an improved global and local solution. The post-refinement multiscale method is assessed using three core designs. When the local solution has more energy groups, the fixed source method has some difficulties near the interface: however the albedo method works well for all cases. In order to remedy the issue with boundary condition errors for the fixed source method, a buffer region is used to act as a filter, which decreases the sensitivity of the solution to the boundary condition. Both the albedo and fixed source methods benefit from the use of a buffer region. Unlike the post-refinement method, the embedded multiscale method alters the global solution. The ability to change the global solution allows for refinement in areas where the errors in the few group nodal diffusion are typically large. The embedded method is shown to improve the global solution when it is applied to a MOX/LEU assembly interface, the fuel/reflector interface, and assemblies where control rods are inserted. The embedded method also allows for multiple solution levels to be applied in a single calculation. The addition of intermediate levels to the solution improves the accuracy of the method. Both multiscale methods considered here have benefits and drawbacks, but both can provide improvements over the current PPR methodology.

EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

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

Matloch, L.; Vaccaro, S.; Couland, M.

The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction ofmore » encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)« less

ADDING REALISM TO NUCLEAR MATERIAL DISSOLVING ANALYSIS

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

Williamson, B.

2011-08-15

Two new criticality modeling approaches have greatly increased the efficiency of dissolver operations in H-Canyon. The first new approach takes credit for the linear, physical distribution of the mass throughout the entire length of the fuel assembly. This distribution of mass is referred to as the linear density. Crediting the linear density of the fuel bundles results in using lower fissile concentrations, which allows higher masses to be charged to the dissolver. Also, this approach takes credit for the fact that only part of the fissile mass is wetted at a time. There are multiple assemblies stacked on top ofmore » each other in a bundle. On average, only 50-75% of the mass (the bottom two or three assemblies) is wetted at a time. This means that only 50-75% (depending on operating level) of the mass is moderated and is contributing to the reactivity of the system. The second new approach takes credit for the progression of the dissolving process. Previously, dissolving analysis looked at a snapshot in time where the same fissile material existed both in the wells and in the bulk solution at the same time. The second new approach models multiple consecutive phases that simulate the fissile material moving from a high concentration in the wells to a low concentration in the bulk solution. This approach is more realistic and allows higher fissile masses to be charged to the dissolver.« less

KSC-08pd0005

NASA Image and Video Library

2008-01-04

KENNEDY SPACE CENTER, FLA. -- At a lab at NASA's Kennedy Space Center, Bob Arp, an aerospace technician with the United Launch Alliance, inserts a wire from an electrical harness onto the pin of a replacement feed-through connector during preparations to solder the pins to the socket of the connector. The connector will be installed in the external fuel tank for space shuttle Atlantis' STS-122 mission. The technician performed this exacting task on the Centaur upper stage for Atlas and Titan launches in 1994 and was specifically chosen for the task. Soldering the connector pins and sockets together addresses the most likely cause of a problem in the engine cutoff sensor system, or ECO system. Some of the tank's ECO sensors failed during propellant tanking for launch attempts on Dec. 6 and Dec. 9. Results of a tanking test on Dec. 18 pointed to an open circuit in the feed-through connector wiring, which is located at the base of the tank. The feed-through connector passes the wires from the inside of the tank to the outside. After the soldering is completed and the connector is reinstalled, shuttle program managers will decide on how to proceed. The launch date for mission STS-122 is under review. Photo credit: NASA/Kim Shiflett

Discrete Ordinate Quadrature Selection for Reactor-based Eigenvalue Problems

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

Jarrell, Joshua J; Evans, Thomas M; Davidson, Gregory G

2013-01-01

In this paper we analyze the effect of various quadrature sets on the eigenvalues of several reactor-based problems, including a two-dimensional (2D) fuel pin, a 2D lattice of fuel pins, and a three-dimensional (3D) reactor core problem. While many quadrature sets have been applied to neutral particle discrete ordinate transport calculations, the Level Symmetric (LS) and the Gauss-Chebyshev product (GC) sets are the most widely used in production-level reactor simulations. Other quadrature sets, such as Quadruple Range (QR) sets, have been shown to be more accurate in shielding applications. In this paper, we compare the LS, GC, QR, and themore » recently developed linear-discontinuous finite element (LDFE) sets, as well as give a brief overview of other proposed quadrature sets. We show that, for a given number of angles, the QR sets are more accurate than the LS and GC in all types of reactor problems analyzed (2D and 3D). We also show that the LDFE sets are more accurate than the LS and GC sets for these problems. We conclude that, for problems where tens to hundreds of quadrature points (directions) per octant are appropriate, QR sets should regularly be used because they have similar integration properties as the LS and GC sets, have no noticeable impact on the speed of convergence of the solution when compared with other quadrature sets, and yield more accurate results. We note that, for very high-order scattering problems, the QR sets exactly integrate fewer angular flux moments over the unit sphere than the GC sets. The effects of those inexact integrations have yet to be analyzed. We also note that the LDFE sets only exactly integrate the zeroth and first angular flux moments. Pin power comparisons and analyses are not included in this paper and are left for future work.« less

Discrete ordinate quadrature selection for reactor-based Eigenvalue problems

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

Jarrell, J. J.; Evans, T. M.; Davidson, G. G.

2013-07-01

In this paper we analyze the effect of various quadrature sets on the eigenvalues of several reactor-based problems, including a two-dimensional (2D) fuel pin, a 2D lattice of fuel pins, and a three-dimensional (3D) reactor core problem. While many quadrature sets have been applied to neutral particle discrete ordinate transport calculations, the Level Symmetric (LS) and the Gauss-Chebyshev product (GC) sets are the most widely used in production-level reactor simulations. Other quadrature sets, such as Quadruple Range (QR) sets, have been shown to be more accurate in shielding applications. In this paper, we compare the LS, GC, QR, and themore » recently developed linear-discontinuous finite element (LDFE) sets, as well as give a brief overview of other proposed quadrature sets. We show that, for a given number of angles, the QR sets are more accurate than the LS and GC in all types of reactor problems analyzed (2D and 3D). We also show that the LDFE sets are more accurate than the LS and GC sets for these problems. We conclude that, for problems where tens to hundreds of quadrature points (directions) per octant are appropriate, QR sets should regularly be used because they have similar integration properties as the LS and GC sets, have no noticeable impact on the speed of convergence of the solution when compared with other quadrature sets, and yield more accurate results. We note that, for very high-order scattering problems, the QR sets exactly integrate fewer angular flux moments over the unit sphere than the GC sets. The effects of those inexact integrations have yet to be analyzed. We also note that the LDFE sets only exactly integrate the zeroth and first angular flux moments. Pin power comparisons and analyses are not included in this paper and are left for future work. (authors)« less

Modeling CANDU-6 liquid zone controllers for effects of thorium-based fuels

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

St-Aubin, E.; Marleau, G.

2012-07-01

We use the DRAGON code to model the CANDU-6 liquid zone controllers and evaluate the effects of thorium-based fuels on their incremental cross sections and reactivity worth. We optimize both the numerical quadrature and spatial discretization for 2D cell models in order to provide accurate fuel properties for 3D liquid zone controller supercell models. We propose a low computer cost parameterized pseudo-exact 3D cluster geometries modeling approach that avoids tracking issues on small external surfaces. This methodology provides consistent incremental cross sections and reactivity worths when the thickness of the buffer region is reduced. When compared with an approximate annularmore » geometry representation of the fuel and coolant region, we observe that the cluster description of fuel bundles in the supercell models does not increase considerably the precision of the results while increasing substantially the CPU time. In addition, this comparison shows that it is imperative to finely describe the liquid zone controller geometry since it has a strong impact of the incremental cross sections. This paper also shows that liquid zone controller reactivity worth is greatly decreased in presence of thorium-based fuels compared to the reference natural uranium fuel, since the fission and the fast to thermal scattering incremental cross sections are higher for the new fuels. (authors)« less

Nuclear Resonance Fluorescence for Materials Assay

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

Quiter, Brian J.; Ludewigt, Bernhard; Mozin, Vladimir

This paper discusses the use of nuclear resonance fluorescence (NRF) techniques for the isotopic and quantitative assaying of radioactive material. Potential applications include age-dating of an unknown radioactive source, pre- and post-detonation nuclear forensics, and safeguards for nuclear fuel cycles Examples of age-dating a strong radioactive source and assaying a spent fuel pin are discussed. The modeling work has ben performed with the Monte Carlo radiation transport computer code MCNPX, and the capability to simulate NRF has bee added to the code. Discussed are the limitations in MCNPX?s photon transport physics for accurately describing photon scattering processes that are importantmore » contributions to the background and impact the applicability of the NRF assay technique.« less

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

NASA Technical Reports Server (NTRS)

Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

1971-01-01

A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7 cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The experimental program consisted basically of measuring the differential neutron spectra and the changes in critical mass that accompanied the stepwise addition of (Li-7)3N, Hf, Ta, and W to a basic core fueled with U metal in a pin-type Ta honeycomb structure. In addition, experimental results were obtained on power distributions, control characteristics, neutron lifetime, and reactivity worths of numerous absorber, structural, and scattering materials.

Pulse magnetic welder

DOEpatents

Christiansen, D.W.; Brown, W.F.

1984-01-01

A welder is described for automated closure of fuel pins by a pulsed magnetic process in which the open end of a length of cladding is positioned within a complementary tube surrounded by a pulsed magnetic welder. Seals are provided at each end of the tube, which can be evacuated or can receive tag gas for direct introduction to the cladding interior. Loading of magnetic rings and end caps is accomplished automatically in conjunction with the welding steps carried out within the tube.

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

Syarifah, Ratna Dewi, E-mail: syarifah.physics@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the additionmore » of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.« less

Self-actuated nuclear reactor shutdown system using induction pump to facilitate sensing of core coolant temperature

DOEpatents

Sievers, Robert K.; Cooper, Martin H.; Tupper, Robert B.

1987-01-01

A self-actuated shutdown system incorporated into a reactivity control assembly in a nuclear reactor includes pumping means for creating an auxiliary downward flow of a portion of the heated coolant exiting from the fuel assemblies disposed adjacent to the control assembly. The shutdown system includes a hollow tubular member which extends through the outlet of the control assembly top nozzle so as to define an outer annular flow channel through the top nozzle outlet separate from an inner flow channel for primary coolant flow through the control assembly. Also, a latching mechanism is disposed in an inner duct of the control assembly and is operable for holding absorber bundles in a raised position in the control assembly and for releasing them to drop them into the core of the reactor for shutdown purposes. The latching mechanism has an inner flow passage extending between and in flow communication with the absorber bundles and the inner flow channel of the top nozzle for accommodating primary coolant flow upwardly through the control assembly. Also, an outer flow passage separate from the inner flow passage extends through the latching mechanism between and in flow communication with the inner duct and the outer flow channel of the top nozzle for accommodating inflow of a portion of the heated coolant from the adjacent fuel assemblies. The latching mechanism contains a magnetic material sensitive to temperature and operable to cause mating or latching together of the components of the latching mechanism when the temperature sensed is below a known temperature and unmating or unlatching thereof when the temperature sensed is above a given temperature. The temperature sensitive magnetic material is positioned in communication with the heated coolant flow through the outer flow passage for directly sensing the temperature thereof. Finally, the pumping means includes a jet induction pump nozzle and diffuser disposed adjacent the bottom nozzle of the control assembly and in flow communication with the inlet thereof. The pump nozzle is operable to create an upward driving flow of primary coolant through the pump diffuser and then to the absorber bundles. The upward driving flow of primary coolant, in turn, creates a suction head within the outer flow channel of the top nozzle and thereby an auxiliary downward flow of the heated coolant portion exiting from the upper end of the adjacent fuel assemblies through the outer flow channel to the pump nozzle via the outer flow passage of the latching mechanism and an annular space between the outer and inner spaced ducts of the control assembly housing. The temperature of the heated coolant exiting from the adjacent fuel assemblies can thereby be sensed directly by the temperature sensitive magnetic material in the latching mechanism.

Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

NASA Astrophysics Data System (ADS)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-01-01

Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

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

Carmack, Jon; Hayes, Steven; Walters, L. C.

This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO 2 and UO 2-PuO 2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availabilitymore » are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.« less

Fuel clad chemical interactions in fast reactor MOX fuels

NASA Astrophysics Data System (ADS)

Viswanathan, R.

2014-01-01

Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel-Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ṡ [B/(at.% fission)] ṡ (T/K-705) ṡ [(O/M)i-1.935]} + 20.5) for (O/M)i ⩽ 1.98. A new model is proposed for (O/M)i ⩾ 1.98: d/μm = [B/(at.% fission)] ṡ (T/K-800)0.5 ṡ [(O/M)i-1.94] ṡ [P/(W cm-1)]0.5. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M)i is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

An evaluation of UO2-CNT composites made by SPS as an accident tolerant nuclear fuel pellet and the feasibility of SPS as an economical fabrication process for the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Cartas, Andrew R.

The innovative and advanced purpose of this study is to understand and establish proper sintering procedures for Spark Plasma Sintering process in order to fabricate high density, high thermal conductivity UO2 -CNT pellets. Mixing quality and chemical reactions have been investigated by field emission scanning electron microscopy (FESEM), wavelength dispersive spectroscopy (WDS), and X-ray diffraction (XRD). The effect of various types of CNTs on the mixing and sintering quality of UO2-CNT pellets with SPS processing have been examined. The Archimedes Immersion Method, laser flash method, and FE-SEM will be used to investigate the density, thermal conductivity, grain size, pinning effects, and CNT dispersion of fabricated UO2-CNT pellets. Pre-fabricated CNT's were added to UO 2 powder and dispersed via sonication and/or ball milling and then made into composite nuclear pellets. An investigation of the economic impact of SPS on the nuclear fuel cycle for producing pure and composite UO2 fuels was conducted.

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

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

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Demonstration of optimum fuel-to-moderator ratio in a PWR unit fuel cell

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

Feltus, M.A.; Pozsgai, C.

1992-01-01

Nuclear engineering students at The Pennsylvania State University develop scaled-down [[approx]350 MW(thermal)] pressurized water reactors (PWRs) using actual plants as references. The design criteria include maintaining the clad temperature below 2200[degree]F, fuel temperature below melting point, sufficient departure from nucleate boiling ratio (DNBR) margin, a beginning-of-life boron concentration that yields a negative moderator temperature coefficient, an adequate cycle power production (330 effective full-power days), and a batch loading scheme that is economical. The design project allows for many degrees of freedom (e.g., assembly number, pitch and height and batch enrichments) so that each student's result is unique. The iterative naturemore » of the design process is stressed in the course. The LEOPARD code is used for the unit cell depletion, critical boron, and equilibrium xenon calculations. Radial two-group diffusion equations are solved with the TWIDDLE-DEE code. The steady-state ZEBRA thermal-hydraulics program is used for calculating DNBR. The unit fuel cell pin radius and pitch (fuel-to-moerator ratio) for the scaled-down design, however, was set equal to the already optimized ratio for the reference PWR. This paper describes an honors project that shows how the optimum fuel-to-moderator ratio is found for a unit fuel cell shown in terms of neutron economics. This exercise illustrates the impact of fuel-to-moderator variations on fuel utilization factor and the effect of assuming space and energy separability.« less

Real-Time Imaging of Self-Organization and Mechanical Competition in Carbon Nanotube Forest Growth.

PubMed

Balakrishnan, Viswanath; Bedewy, Mostafa; Meshot, Eric R; Pattinson, Sebastian W; Polsen, Erik S; Laye, Fabrice; Zakharov, Dmitri N; Stach, Eric A; Hart, A John

2016-12-27

The properties of carbon nanotube (CNT) networks and analogous materials comprising filamentary nanostructures are governed by the intrinsic filament properties and their hierarchical organization and interconnection. As a result, direct knowledge of the collective dynamics of CNT synthesis and self-organization is essential to engineering improved CNT materials for applications such as membranes and thermal interfaces. Here, we use real-time environmental transmission electron microscopy (E-TEM) to observe nucleation and self-organization of CNTs into vertically aligned forests. Upon introduction of the carbon source, we observe a large scatter in the onset of nucleation of individual CNTs and the ensuing growth rates. Experiments performed at different temperatures and catalyst particle densities show the critical role of CNT density on the dynamics of self-organization; low-density CNT nucleation results in the CNTs becoming pinned to the substrate and forming random networks, whereas higher density CNT nucleation results in self-organization of the CNTs into bundles that are oriented perpendicular to the substrate. We also find that mechanical coupling between growing CNTs alters their growth trajectory and shape, causing significant deformations, buckling, and defects in the CNT walls. Therefore, it appears that CNT-CNT coupling not only is critical for self-organization but also directly influences CNT quality and likely the resulting properties of the forest. Our findings show that control of the time-distributed kinetics of CNT nucleation and bundle formation are critical to manufacturing well-organized CNT assemblies and that E-TEM can be a powerful tool to investigate the mesoscale dynamics of CNT networks.

Tribological evaluation of an Al2O3-SiO2 ceramic fiber candidate for high-temperature sliding seals

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

Dellacorte, C.; Steinetz, B.

A test program to determine the relative sliding durability of an alumina-silica candidate ceramic fiber for high temperature sliding seal applications as described. This work represents the first reporting of the sliding durability of this material system. Pin-on-disk tests were used to evaluate the potential seal material by sliding a tow or bundle of the candidate ceramic fiber against a superalloy test disk. Friction was measured during the tests and fiber wear, indicated by the extent of fibers broken in the tow or bundle, was measured at the end of each test. Test variables studied included ambient temperatures from 25more » C to 900 C, loads from 1.3 to 21.2 Newtons, and sliding velocities from 0.025 to 0.25 m/sec. In addition, the effects of fiber diameter, elastic modulus, and a pretest fiber heat treatment on friction and wear were measured. In most cases, wear increased with temperature. Friction ranged from about 0.36 at 500 C and low velocity (0.025 m/s) to over 1.1 at 900 C and high velocity (0.25 m/s). The pretest fiber heat treatment, which caused significant durability reductions for alumina-boria-silica ceramic fibers tested previously, had little effect on the alumina-silica fibers tested here. These results indicate that the alumina-silica (Al2O3-SiO2) fiber is a good candidate material system for high temperature sliding seal applications.« less

The Interdependence of Various Types of Autoignition and Knock

NASA Technical Reports Server (NTRS)

Olsen, H Lowell; Miller, Cearcy D

1948-01-01

A study of the relations existing among pin-point autoignition, homogeneous autoignition, and knock has been made by means of the NACA high-speed camera and the full-view combustion apparatus. High-speed photographic records of combustion, together with corresponding pressure-time traces, of benzene, 2,2,3-trimethylbutane, S-4, and M-4 fuels at various engine conditions have shown the engine conditions under which each of these phenomena occur and the relation of these phenomena to one another.

Sliding wear and friction behaviour of zircaloy-4 in water

NASA Astrophysics Data System (ADS)

Sharma, Garima; Limaye, P. K.; Jadhav, D. T.

2009-11-01

In water cooled nuclear reactors, the sliding of fuel bundles in fuel channel handling system can lead to severe wear and it is an important topic to study. In the present study, sliding wear behaviour of zircaloy-4 was investigated in water (pH ˜ 10.5) using ball-on-plate sliding wear tester. Sliding wear resistance zircaloy-4 against SS 316 was examined at room temperature. Sliding wear tests were carried out at different load and sliding frequencies. The coefficient of friction of zircaloy-4 was also measured during each tests and it was found to decrease slightly with the increase in applied load. The micro-mechanisms responsible for wear in zircaloy-4 were identified to be microcutting, micropitting and microcracking of deformed subsurface zones in water.

X-ray coherent diffraction imaging of cellulose fibrils in situ.

PubMed

Lal, Jyotsana; Harder, Ross; Makowski, Lee

2011-01-01

Cellulose is the most abundant renewable source of organic molecules on earth[1]. As fossil fuel reserves become depleted, the use of cellulose as a feed stock for fuels and chemicals is being aggressively explored. Cellulose is a linear polymer of glucose that packs tightly into crystalline fibrils that make up a substantial proportion of plant cell walls. Extraction of the cellulose chains from these fibrils in a chemically benign process has proven to be a substantial challenge [2]. Monitoring the deconstruction of the fibrils in response to physical and chemical treatments would expedite the development of efficient processing methods. As a step towards achieving that goal, we here describe Bragg-coherent diffraction imaging (CDI) as an approach to producing images of cellulose fibrils in situ within vascular bundles from maize.

Superconducting RF Linacs Driving Subcritical Reactors for Profitable Disposition of Surplus Weapons-grade Plutonium

NASA Astrophysics Data System (ADS)

Cummings, Mary Anne; Johnson, Rolland

Acceptable capital and operating costs of high-power proton accelerators suitable for profitable commercial electric-power and process-heat applications have been demonstrated. However, studies have pointed out that even a few hundred trips of an accelerator lasting a few seconds would lead to unacceptable thermal stresses as each trip causes fission to be turned off in solid fuel structures found in conventional reactors. The newest designs based on the GEM*STAR concept take such trips in stride by using molten-salt fuel, where fuel pin fatigue is not an issue. Other aspects of the GEM*STAR concept which address all historical reactor failures include an internal spallation neutron target and high temperature molten salt fuel with continuous purging of volatile radioactive fission products such that the reactor contains less than a critical mass and almost a million times fewer volatile radioactive fission products than conventional reactors. GEM*STAR is a reactor that without redesign will burn spent nuclear fuel, natural uranium, thorium, or surplus weapons material. It will operate without the need for a critical core, fuel enrichment, or reprocessing making it an excellent candidate for export. As a first application, the design for a pilot plant is described for the profitable disposition of surplus weapons-grade plutonium by using process heat to produce green diesel fuel for the Department of Defense (DOD) from natural gas and renewable carbon.

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

PubMed Central

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

2016-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Safety and Regulatory Issues of the Thorium Fuel Cycle

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

Ade, Brian; Worrall, Andrew; Powers, Jeffrey

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2),more » add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.« less

A 100-kWt NaK-Cooled Space Reactor Concept for an Early-Flight Mission

NASA Astrophysics Data System (ADS)

Poston, David I.

2003-01-01

A stainless-steel (SS) sodium-potassium (NaK) cooled reactor could potentially be the first step in utilizing fission technology in space. The sum of all system-level experience for liquid-metal-cooled space reactors has been with NaK, including the SNAP-10a, the only reactor ever launched by the US. This paper describes a 100-kWt NaK reactor, the NaK-100, which is designed to be developed with minimal technical risk. In additional to NaK technology heritage, the NaK-100 uses a proven fuel-form (SS/UO2) and is designed for simplified system integration and testing. The pins are placed within a solid SS prism, and the NaK flows in an annulus between the pins and the prism. The nuclear and thermal-hydraulic performance of the NaK-100 is presented, as well as the major differences between the NaK-100 and SNAP-10a.

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

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

Aly, A.; Avramova, Maria; Ivanov, Kostadin

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

Flight Testing of Terrain-Relative Navigation and Large-Divert Guidance on a VTVL Rocket

NASA Technical Reports Server (NTRS)

Trawny, Nikolas; Benito, Joel; Tweddle, Brent; Bergh, Charles F.; Khanoyan, Garen; Vaughan, Geoffrey M.; Zheng, Jason X.; Villalpando, Carlos Y.; Cheng, Yang; Scharf, Daniel P.;

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

DOE PAGES

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

2016-10-01

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

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

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

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

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

The JRC-ITU approach to the safety of advanced nuclear fuel cycles

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

Fanghaenel, T.; Rondinella, V.V.; Somers, J.

2013-07-01

The JRC-ITU safety studies of advanced fuels and cycles adopt two main axes. First the full exploitation of still available and highly relevant knowledge and samples from past fuel preparation and irradiation campaigns (complementing the limited number of ongoing programmes). Secondly, the shift of focus from simple property measurement towards the understanding of basic mechanisms determining property evolution and behaviour of fuel compounds during normal, off-normal and accident conditions. The final objective of the second axis is the determination of predictive tools applicable to systems and conditions different from those from which they were derived. State of the art experimentalmore » facilities, extensive networks of partnerships and collaboration with other organizations worldwide, and a developing programme for training and education are essential in this approach. This strategy has been implemented through various programs and projects. The SUPERFACT programme constitutes the main body of existing knowledge on the behavior in-pile of MOX fuel containing minor actinides. It encompassed all steps of a closed fuel cycle. Another international project investigating the safety of a closed cycle is METAPHIX. In this case a U-Pu19-Zr10 metal alloy containing Np, Am and Cm constitutes the fuel. 9 test pins have been prepared and irradiated. In addition to the PIE (Post Irradiation Examination), pyrometallurgical separation of the irradiated fuel has been performed, to demonstrate all the steps of a multiple recycling closed cycle and characterize their safety relevant aspects. Basic studies like thermodynamic fuel properties, fuel-cladding-coolant interactions have also been carried out at JRC-ITU.« less

Photoreduction of CO{sub 2} by TiO{sub 2} nanocomposites synthesized through reactive direct current magnetron sputter deposition.

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

Chen, L.; Graham, M. E.; Li, G.

The photoreduction of CO{sub 2} into methane provides a carbon-neutral energy alternative to fossil fuels, but its feasibility requires improvements in the photo-efficiency of materials tailored to this reaction. We hypothesize that mixed phase TiO{sub 2} nano-materials with high interfacial densities are extremely active photocatalysts well suited to solar fuel production by reducing CO{sub 2} to methane and shifting to visible light response. Mixed phase TiO{sub 2} films were synthesized by direct current (DC) magnetron sputtering and characterized by X-ray diffraction (XRD), atomic force microscope (AFM), scanning electron microscope (SEM) and transmission electron microscope (TEM). Bundles of anatase-rutile nano-columns havingmore » high densities of two kinds of interfaces (those among the bundles and those between the columns) are fabricated. Films sputtered at a low deposition angle showed the highest methane yield, compared to TiO{sub 2} fabricated under other sputtering conditions and commercial standard Degussa P25 under UV irradiation. The yield of methane could be significantly increased ({approx} 12% CO{sub 2} conversion) by increasing the CO{sub 2} to water ratio and temperature (< 100 C) as a combined effect. These films also displayed a light response strongly shifted into the visible range. This is explained by the creation of non-stoichiometric titania films having unique features that we can potentially tailor to the solar energy applications.« less

Deterministic Local Sensitivity Analysis of Augmented Systems - II: Applications to the QUENCH-04 Experiment Using the RELAP5/MOD3.2 Code System

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

Ionescu-Bujor, Mihaela; Jin Xuezhou; Cacuci, Dan G.

2005-09-15

The adjoint sensitivity analysis procedure for augmented systems for application to the RELAP5/MOD3.2 code system is illustrated. Specifically, the adjoint sensitivity model corresponding to the heat structure models in RELAP5/MOD3.2 is derived and subsequently augmented to the two-fluid adjoint sensitivity model (ASM-REL/TF). The end product, called ASM-REL/TFH, comprises the complete adjoint sensitivity model for the coupled fluid dynamics/heat structure packages of the large-scale simulation code RELAP5/MOD3.2. The ASM-REL/TFH model is validated by computing sensitivities to the initial conditions for various time-dependent temperatures in the test bundle of the Quench-04 reactor safety experiment. This experiment simulates the reflooding with water ofmore » uncovered, degraded fuel rods, clad with material (Zircaloy-4) that has the same composition and size as that used in typical pressurized water reactors. The most important response for the Quench-04 experiment is the time evolution of the cladding temperature of heated fuel rods. The ASM-REL/TFH model is subsequently used to perform an illustrative sensitivity analysis of this and other time-dependent temperatures within the bundle. The results computed by using the augmented adjoint sensitivity system, ASM-REL/TFH, highlight the reliability, efficiency, and usefulness of the adjoint sensitivity analysis procedure for computing time-dependent sensitivities.« less

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

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

Swinhoe, Martyn Thomas; De Baere, Paul

2015-04-17

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

SP-100 multimegawatt scaleup to meet electric propulsion mission requirements

NASA Astrophysics Data System (ADS)

Newkirk, D. W.; Salamah, S. A.; Stewart, S. L.; Pluta, P. R.

The SP-100 nuclear heat source technology, utilizing uranium nitride fuel clad in PWC-11 in a fast reactor with lithium coolant circulated by an electromagnetic pump, is shown to be directly extrapolatable to thermal power levels that meet NASA nuclear electric propulsion requirements using different power conversion techniques. The SP-100 nuclear technology can be applied to missions with NEP (nuclear electric propulsion) requirements as low as tens of kWe to tens of MWe. It is pointed out that the SP-100 heat source has a great advantage of very long lifetime capability, since it utilizes very rugged refractory metal fuel pins and is independent of the power conversion scheme chosen for a given mission. The only moving parts in the nuclear subsystems are the control rods moved to compensate for fuel enrichment degradation due to fission and for power shutdown. Lowest alpha values in the range of interest for potential NASA missions are predicted for the dynamic Rankine and static HYTEC conversion systems.

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

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

Craft, Aaron E.; Papaioannou, Glen C.; Chichester, David L.

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutronmore » radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.« less

Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

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

None

The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution ofmore » 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments with the Advanced TREAT loop, and the stress placed on these calculations reflects this. Additional material on performance and safety is provided. Backup calculations on calculations of plutonium-release limits are described. Cost and schedule information for the Upgrade are presented.« less

Diagnosing Undersampling in Monte Carlo Eigenvalue and Flux Tally Estimates

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

Perfetti, Christopher M; Rearden, Bradley T

2015-01-01

This study explored the impact of undersampling on the accuracy of tally estimates in Monte Carlo (MC) calculations. Steady-state MC simulations were performed for models of several critical systems with varying degrees of spatial and isotopic complexity, and the impact of undersampling on eigenvalue and fuel pin flux/fission estimates was examined. This study observed biases in MC eigenvalue estimates as large as several percent and biases in fuel pin flux/fission tally estimates that exceeded tens, and in some cases hundreds, of percent. This study also investigated five statistical metrics for predicting the occurrence of undersampling biases in MC simulations. Threemore » of the metrics (the Heidelberger-Welch RHW, the Geweke Z-Score, and the Gelman-Rubin diagnostics) are commonly used for diagnosing the convergence of Markov chains, and two of the methods (the Contributing Particles per Generation and Tally Entropy) are new convergence metrics developed in the course of this study. These metrics were implemented in the KENO MC code within the SCALE code system and were evaluated for their reliability at predicting the onset and magnitude of undersampling biases in MC eigenvalue and flux tally estimates in two of the critical models. Of the five methods investigated, the Heidelberger-Welch RHW, the Gelman-Rubin diagnostics, and Tally Entropy produced test metrics that correlated strongly to the size of the observed undersampling biases, indicating their potential to effectively predict the size and prevalence of undersampling biases in MC simulations.« less

The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

NASA Astrophysics Data System (ADS)

Syarifah, Ratna Dewi; Suud, Zaki

2015-09-01

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

Design of pellet surface grooves for fission gas plenum

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

Carter, T.J.; Jones, L.R.; Macici, N.

1986-01-01

In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the methodology for design of gas storage volume within the fuel pellets - specifically the use of grooves pressed or machined into the relatively cool pellet/cladding interface. Preanalysis and design of pellet groove shape and volume was accomplished using the TRUMPmore » heat transfer code. Postirradiation examination (PIE) was used to check the initial design and heat transfer assumptions. Fission gas release was found to be higher for the grooved pellet rods than for the comparison rods with hollow or unmodified pellets. This had been expected from the initial TRUMP thermal analyses. The ELESIM fuel modeling code was used to check in-reactor performance, but some modifications were necessary to accommodate the loss of heat transfer surface to the grooves. It was concluded that for plenum design purposes, circumferential pellet grooves could be adequately modeled by the codes TRUMP and ELESIM.« less

Improving Prompt Temperature Feedback by Stimulating Doppler Broadening in Heterogeneous Composite Nuclear Fuel Forms

NASA Astrophysics Data System (ADS)

Morrison, Christopher

Nuclear fuels with similar aggregate material composition, but with different millimeter and micrometer spatial configurations of the component materials can have very different safety and performance characteristics. This research focuses on modeling and attempting to engineer heterogeneous combinations of nuclear fuels to improve negative prompt temperature feedback in response to reactivity insertion accidents. Improvements in negative prompt temperature feedback are proposed by developing a tailored thermal resistance in the nuclear fuel. In the event of a large reactivity insertion, the thermal resistance allows for a faster negative Doppler feedback by temporarily trapping heat in material zones with strong absorption resonances. A multi-physics simulation framework was created that could model large reactivity insertions. The framework was then used to model a comparison of a heterogeneous fuel with a tailored thermal resistance and a homogeneous fuel without the tailored thermal resistance. The results from the analysis confirmed the fundamental premise of prompt temperature feedback and provide insights into the neutron spectrum dynamics throughout the transient process. A trade study was conducted on infinite lattice fuels to help map a design space to study and improve prompt temperature feedback with many results. A multi-scale fuel pin analysis was also completed to study more realistic geometries. The results of this research could someday allow for novel nuclear fuels that would behave differently than current fuels. The idea of having a thermal barrier coating in the fuel is contrary to most current thinking. Inherent resistance to reactivity insertion accidents could enable certain reactor types once considered vulnerable to reactivity insertion accidents to be reevaluated in light of improved negative prompt temperature feedback.

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

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

Daniel, G.; Rudisill, T.; Almond, P.

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

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

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

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

The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

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

Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier

The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimetermore » scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.« less

In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

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

Burgett, Eric

2015-10-13

An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In additionmore » to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.« less

Refining of Military Jet Fuels from Shale Oil. Part II. Volume II. (In Situ Shale Oil Process Data).

DTIC Science & Technology

1982-03-01

SPEC Meeting Specifications OXY Test Series on In Situ Shale Oil z P Pressure (P + N) Paraffins and Naphthenes PRO Test Series on Above Ground Shale Oil...LV 6/ 12.0 Naphthenes , LV% (Aromatics, LV %/ 11.8 Gross Heating Value, Btu/lb 19,720 19,068 -73- TABLE 111-29. CRUDE SHALE: OIL HYDROTREATING SERIES M...Wt % - Ramabottomn Carbon -1.34 IParaffins (P-IN), LV % (71.1) -IOlef ins, LV % 9.4 i ~ Naphthenes , LV% - Aromatics, LV % 19.5 - Gross Heating Value

TREAT neutron-radiography facility

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

Harrison, L.J.

1981-01-01

The TREAT reactor was built as a transient irradiation test reactor. By taking advantage of built-in system features, it was possible to add a neutron-radiography facility. This facility has been used over the years to radiograph a wide variety and large number of preirradiated fuel pins in many different configurations. Eight different specimen handling casks weighing up to 54.4 t (60 T) can be accommodated. Thermal, epithermal, and track-etch radiographs have been taken. Neutron-radiography service can be provided for specimens from other reactor facilities, and the capacity for storing preirradiated specimens also exists.

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

Performance tests of a single-cylinder compression-ignition engine with a displacer piston

NASA Technical Reports Server (NTRS)

Moore, C S; Foster, H H

1935-01-01

Engine performance was investigated using a rectangular displacer on the piston crown to cause a forced air flow in a vertical-disk combustion chamber of a single-cylinder, 4-stroke-cycle compression-ignition engine. The optimum air-flow area was determined first with the area concentrated at one end of the displacer and then with the area equally divided between two passages, one at each end of the displacer. Best performance was obtained with the two-passage air flow arranged to give a calculated maximum air-flow speed of 8 times the linear crank-pin speed. With the same fuel-spray formation as used without the air flow, the maximum clear exhaust brake mean effective pressure at 1,500 r.p.m. was increased from 90 to 115 pounds per square inch and the corresponding fuel consumption reduced from 0.46 to 0.43 pound per brake horsepower-hour. At 1,200 r.p.m., a maximum clear exhaust brake mean effective pressure of 120 pounds per square inch was obtained at a fuel consumption of 0.42 pound per brake horsepower-hour. At higher specific fuel consumption the brake mean effective pressure was still increasing rapidly.

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

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

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less

A User Guide to PARET/ANL

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

Olson, A. P.; Dionne, B.; Marin-Lafleche, A.

2015-01-01

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA, and has been used by the Reactor Conversion Program tomore » determine the expected transient behavior of a large number of reactors. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.« less

Real-time imaging of self-organization and mechanical competition in carbon nanotube forest growth [Real-time environmental TEM investigation of self-organization and mechanical competition in carbon nanotube forest growth

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

Balakrishnan, Viswanath; Bedewy, Mostafa; Meshot, Eric R.

Here, the properties of carbon nanotube (CNT) networks and analogous materials comprising filamentary nanostructures are governed by the intrinsic filament properties and their hierarchical organization and interconnection. As a result, direct knowledge of the collective dynamics of CNT synthesis and self-organization is essential to engineering improved CNT materials for applications such as membranes and thermal interfaces. Here, we use real-time environmental transmission electron microscopy (E-TEM) to observe nucleation and self-organization of CNTs into vertically aligned forests. Upon introduction of the carbon source, we observe a large scatter in the onset of nucleation of individual CNTs and the ensuing growth rates.more » Experiments performed at different temperatures and catalyst particle densities show the critical role of CNT density on the dynamics of self-organization; low-density CNT nucleation results in the CNTs becoming pinned to the substrate and forming random networks, whereas higher density CNT nucleation results in self-organization of the CNTs into bundles that are oriented perpendicular to the substrate. We also find that mechanical coupling between growing CNTs alters their growth trajectory and shape, causing significant deformations, buckling, and defects in the CNT walls. Therefore, it appears that CNT–CNT coupling not only is critical for self-organization but also directly influences CNT quality and likely the resulting properties of the forest. As a result, our findings show that control of the time-distributed kinetics of CNT nucleation and bundle formation are critical to manufacturing well-organized CNT assemblies and that E-TEM can be a powerful tool to investigate the mesoscale dynamics of CNT networks.« less

Real-time imaging of self-organization and mechanical competition in carbon nanotube forest growth [Real-time environmental TEM investigation of self-organization and mechanical competition in carbon nanotube forest growth

DOE PAGES

Balakrishnan, Viswanath; Bedewy, Mostafa; Meshot, Eric R.; ...

2016-11-23

Here, the properties of carbon nanotube (CNT) networks and analogous materials comprising filamentary nanostructures are governed by the intrinsic filament properties and their hierarchical organization and interconnection. As a result, direct knowledge of the collective dynamics of CNT synthesis and self-organization is essential to engineering improved CNT materials for applications such as membranes and thermal interfaces. Here, we use real-time environmental transmission electron microscopy (E-TEM) to observe nucleation and self-organization of CNTs into vertically aligned forests. Upon introduction of the carbon source, we observe a large scatter in the onset of nucleation of individual CNTs and the ensuing growth rates.more » Experiments performed at different temperatures and catalyst particle densities show the critical role of CNT density on the dynamics of self-organization; low-density CNT nucleation results in the CNTs becoming pinned to the substrate and forming random networks, whereas higher density CNT nucleation results in self-organization of the CNTs into bundles that are oriented perpendicular to the substrate. We also find that mechanical coupling between growing CNTs alters their growth trajectory and shape, causing significant deformations, buckling, and defects in the CNT walls. Therefore, it appears that CNT–CNT coupling not only is critical for self-organization but also directly influences CNT quality and likely the resulting properties of the forest. As a result, our findings show that control of the time-distributed kinetics of CNT nucleation and bundle formation are critical to manufacturing well-organized CNT assemblies and that E-TEM can be a powerful tool to investigate the mesoscale dynamics of CNT networks.« less

Patterning of leaf vein networks by convergent auxin transport pathways.

PubMed

Sawchuk, Megan G; Edgar, Alexander; Scarpella, Enrico

2013-01-01

The formation of leaf vein patterns has fascinated biologists for centuries. Transport of the plant signal auxin has long been implicated in vein patterning, but molecular details have remained unclear. Varied evidence suggests a central role for the plasma-membrane (PM)-localized PIN-FORMED1 (PIN1) intercellular auxin transporter of Arabidopsis thaliana in auxin-transport-dependent vein patterning. However, in contrast to the severe vein-pattern defects induced by auxin transport inhibitors, pin1 mutant leaves have only mild vein-pattern defects. These defects have been interpreted as evidence of redundancy between PIN1 and the other four PM-localized PIN proteins in vein patterning, redundancy that underlies many developmental processes. By contrast, we show here that vein patterning in the Arabidopsis leaf is controlled by two distinct and convergent auxin-transport pathways: intercellular auxin transport mediated by PM-localized PIN1 and intracellular auxin transport mediated by the evolutionarily older, endoplasmic-reticulum-localized PIN6, PIN8, and PIN5. PIN6 and PIN8 are expressed, as PIN1 and PIN5, at sites of vein formation. pin6 synthetically enhances pin1 vein-pattern defects, and pin8 quantitatively enhances pin1pin6 vein-pattern defects. Function of PIN6 is necessary, redundantly with that of PIN8, and sufficient to control auxin response levels, PIN1 expression, and vein network formation; and the vein pattern defects induced by ectopic PIN6 expression are mimicked by ectopic PIN8 expression. Finally, vein patterning functions of PIN6 and PIN8 are antagonized by PIN5 function. Our data define a new level of control of vein patterning, one with repercussions on other patterning processes in the plant, and suggest a mechanism to select cell files specialized for vascular function that predates evolution of PM-localized PIN proteins.

Burnable absorber arrangement for fuel bundle

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

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

1986-12-16

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

Development and Assessment of CTF for Pin-resolved BWR Modeling

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

Salko, Robert K; Wysocki, Aaron J; Collins, Benjamin S

2017-01-01

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CSmore » workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.« less

Portal placement in elbow arthroscopy by novice surgeons: cadaver study.

PubMed

Claessen, Femke M A P; Kachooei, Amir R; Kolovich, Gregory P; Buijze, Geert A; Oh, Luke S; van den Bekerom, Michel P J; Doornberg, Job N

2017-07-01

In this anatomical cadaver study, the distance between major nerves and ligaments at risk for injury and portal sites created by trainees was measured. Trainees, inexperienced in elbow arthroscopy, have received a didactic lecture and cadaver instruction prior to portal placement. The incidence of iatrogenic injury from novice portal placement was also determined. Anterolateral, direct lateral, and anteromedial arthroscopic portals were created in ten cadavers by ten inexperienced trainees in elbow arthroscopy. After creating each portal, the trajectory of the portal was marked with a guide pin. Subsequently, the cadavers were dissected and the distances between the guide pin in the anterolateral, direct lateral, and anteromedial portals and important ligaments and nerves were measured. The difference between the distance of the direct lateral portal and the posterior antebrachial cutaneous nerve (PABCN) (22 mm, p < 0.001), the lateral antebrachial cutaneous nerve (4.0 mm, p < 0.001), and the radial nerve (25 mm, p < 0.001) was different from the average reported distances in the literature. A difference was found between the distance of the anterolateral portal and the PABCN (32 mm, p < 0.001) compared to previous studies. Three major iatrogenic complications were observed, including: laceration of the posterior bundle of the medial ulnar collateral ligament, lateral ulnar collateral ligament midsubstance laceration, and median nerve partial laceration. Surgeons increasingly consider arthroscopic treatment as an option for elbow pathology. In the present study a surgical complication rate of 30 % was found with novice portal placement during elbow arthroscopy. Furthermore, as the results from this study have indicated, accurate, precise, and safe portal placement in elbow arthroscopy is not easily achieved by didactic lecture and cadaver instruction session alone. Level of evidence V.

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

NASA Astrophysics Data System (ADS)

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

2012-07-01

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

Cross-species functional diversity within the PIN auxin efflux protein family

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

O'Connor, Devin Lee; Elton, Samuel; Ticchiarelli, Fabrizio

In Arabidopsis, development during flowering is coordinated by transport of the hormone auxin mediated by polar-localized PIN-FORMED1 (AtPIN1). However Arabidopsis has lost a PIN clade sister to AtPIN1, Sister-of-PIN1 (SoPIN1), which is conserved in flowering plants. Here, we previously proposed that the AtPIN1 organ initiation and vein patterning functions are split between the SoPIN1 and PIN1 clades in grasses. Here we show that in the grass Brachypodium sopin1 mutants have organ initiation defects similar to Arabidopsis atpin1, while loss of PIN1 function in Brachypodium has little effect on organ initiation but alters stem growth. Heterologous expression of Brachypodium SoPIN1 andmore » PIN1b in Arabidopsis provides further evidence of functional specificity. SoPIN1 but not PIN1b can mediate flower formation in null atpin1 mutants, although both can complement a missense allele. The behavior of SoPIN1 and PIN1b in Arabidopsis illustrates how membrane and tissue-level accumulation, transport activity, and interaction contribute to PIN functional specificity.« less

Cross-species functional diversity within the PIN auxin efflux protein family

DOE PAGES

O'Connor, Devin Lee; Elton, Samuel; Ticchiarelli, Fabrizio; ...

2017-10-24

In Arabidopsis, development during flowering is coordinated by transport of the hormone auxin mediated by polar-localized PIN-FORMED1 (AtPIN1). However Arabidopsis has lost a PIN clade sister to AtPIN1, Sister-of-PIN1 (SoPIN1), which is conserved in flowering plants. Here, we previously proposed that the AtPIN1 organ initiation and vein patterning functions are split between the SoPIN1 and PIN1 clades in grasses. Here we show that in the grass Brachypodium sopin1 mutants have organ initiation defects similar to Arabidopsis atpin1, while loss of PIN1 function in Brachypodium has little effect on organ initiation but alters stem growth. Heterologous expression of Brachypodium SoPIN1 andmore » PIN1b in Arabidopsis provides further evidence of functional specificity. SoPIN1 but not PIN1b can mediate flower formation in null atpin1 mutants, although both can complement a missense allele. The behavior of SoPIN1 and PIN1b in Arabidopsis illustrates how membrane and tissue-level accumulation, transport activity, and interaction contribute to PIN functional specificity.« less

Pin1At regulates PIN1 polar localization and root gravitropism.

PubMed

Xi, Wanyan; Gong, Ximing; Yang, Qiaoyun; Yu, Hao; Liou, Yih-Cherng

2016-01-21

Root gravitropism allows plants to establish root systems and its regulation depends on polar auxin transport mediated by PIN-FORMED (PIN) auxin transporters. PINOID (PID) and PROTEIN PHOSPHATASE 2A (PP2A) act antagonistically on reversible phosphorylation of PINs. This regulates polar PIN distribution and auxin transport. Here we show that a peptidyl-prolyl cis/trans isomerase Pin1At regulates root gravitropism. Downregulation of Pin1At suppresses root agravitropic phenotypes of pp2aa and 35S:PID, while overexpression of Pin1At affects root gravitropic responses and enhances the pp2aa agravitropic phenotype. Pin1At also affects auxin transport and polar localization of PIN1 in stele cells, which is mediated by PID and PP2A. Furthermore, Pin1At catalyses the conformational change of the phosphorylated Ser/Thr-Pro motifs of PIN1. Thus, Pin1At mediates the conformational dynamics of PIN1 and affects PID- and PP2A-mediated regulation of PIN1 polar localization, which correlates with the regulation of root gravitropism.

Pin1At regulates PIN1 polar localization and root gravitropism

PubMed Central

Xi, Wanyan; Gong, Ximing; Yang, Qiaoyun; Yu, Hao; Liou, Yih-Cherng

2016-01-01

Root gravitropism allows plants to establish root systems and its regulation depends on polar auxin transport mediated by PIN-FORMED (PIN) auxin transporters. PINOID (PID) and PROTEIN PHOSPHATASE 2A (PP2A) act antagonistically on reversible phosphorylation of PINs. This regulates polar PIN distribution and auxin transport. Here we show that a peptidyl-prolyl cis/trans isomerase Pin1At regulates root gravitropism. Downregulation of Pin1At suppresses root agravitropic phenotypes of pp2aa and 35S:PID, while overexpression of Pin1At affects root gravitropic responses and enhances the pp2aa agravitropic phenotype. Pin1At also affects auxin transport and polar localization of PIN1 in stele cells, which is mediated by PID and PP2A. Furthermore, Pin1At catalyses the conformational change of the phosphorylated Ser/Thr-Pro motifs of PIN1. Thus, Pin1At mediates the conformational dynamics of PIN1 and affects PID- and PP2A-mediated regulation of PIN1 polar localization, which correlates with the regulation of root gravitropism. PMID:26791759

A numerical study on liquid charging inside electrostatic atomizers

NASA Astrophysics Data System (ADS)

Kashir, Babak; Perri, Anthony; Sankaran, Abhilash; Staszel, Christopher; Yarin, Alexander; Mashayek, Farzad

2016-11-01

The charging of the dielectric liquid inside an electrostatic atomizer is studied numerically by developing codes based on the OpenFOAM platform. Electrostatic atomization is an appealing technology in painting, fuel injection and oil coating systems due to improved particle-size distribution, enhanced controlability of droplets' trajectories and lower power consumption. The numerical study is conducted concurrently to an experimental investigation to facilitate the validation and deliver feedback for further development. The atomizer includes a pin electrode that is placed at the center of a converging chamber. The chamber orifice is located at a known distance from the electrode tip. The pin electrode is connected to a high voltage that leads to the charging of the liquid. In the present work, the theoretical foundations of separated treatment of the polarized layer and the electronuetral bulk flow are set by describing the governing equations, relevant boundary conditions and the matching condition between these two domains. The resulting split domains are solved numerically to find the distribution of velocity and electrostatic fields over the specified regions. National Science Foundation Award Number: 1505276.

KSC-08pd0008

NASA Image and Video Library

2008-01-04

KENNEDY SPACE CENTER, FLA. -- At a lab at NASA's Kennedy Space Center, the first two of nine pins have been soldered to the socket of the replacement feed-through connector that will be installed in the external fuel tank for space shuttle Atlantis' STS-122 mission. Two United Launch Alliance technicians, who performed this exacting task on the Centaur upper stage for Atlas and Titan launches in 1994, will be doing the soldering. Soldering the connector pins and sockets together addresses the most likely cause of a problem in the engine cutoff sensor system, or ECO system. Some of the tank's ECO sensors failed during propellant tanking for launch attempts on Dec. 6 and Dec. 9. Results of a tanking test on Dec. 18 pointed to an open circuit in the feed-through connector wiring, which is located at the base of the tank. The feed-through connector passes the wires from the inside of the tank to the outside. After the soldering is completed and the connector is reinstalled, shuttle program managers will decide on how to proceed. The launch date for mission STS-122 is under review. Photo credit: NASA/Kim Shiflett

KSC-08pd0007

NASA Image and Video Library

2008-01-04

KENNEDY SPACE CENTER, FLA. -- At a lab at NASA's Kennedy Space Center, the first two of nine pins have been soldered to the socket of the replacement feed-through connector that will be installed in the external fuel tank for space shuttle Atlantis' STS-122 mission. Two United Launch Alliance technicians, who performed this exacting task on the Centaur upper stage for Atlas and Titan launches in 1994, will be doing the soldering. Soldering the connector pins and sockets together addresses the most likely cause of a problem in the engine cutoff sensor system, or ECO system. Some of the tank's ECO sensors failed during propellant tanking for launch attempts on Dec. 6 and Dec. 9. Results of a tanking test on Dec. 18 pointed to an open circuit in the feed-through connector wiring, which is located at the base of the tank. The feed-through connector passes the wires from the inside of the tank to the outside. After the soldering is completed and the connector is reinstalled, shuttle program managers will decide on how to proceed. The launch date for mission STS-122 is under review. Photo credit: NASA/Kim Shiflett

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın

2011-07-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other hand, in order to accommodate solid fission product swelling and to control fuel clad mechanical interaction of the stiffer fuel, the fuel smear density is reduced to 70%. In addition, plenum height is increased to accommodate for fission gases.

Model calculation of Cr dissolution behavior of ODS ferritic steel in high-temperature flowing sodium environment

NASA Astrophysics Data System (ADS)

Ohtsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji

2018-07-01

A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in inlet sodium bulk flow was set at 0.07 wt ppm in the calculation.

Coupled field-structural analysis of HGTR fuel brick using ABAQUS

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

Mohanty, S.; Jain, R.; Majumdar, S.

2012-07-01

High-temperature, gas-cooled reactors (HTGRs) are usually helium-gas cooled, with a graphite core that can operate at reactor outlet temperatures much higher than can conventional light water reactors. In HTGRs, graphite components moderate and reflect neutrons. During reactor operation, high temperature and high irradiation cause damage to the graphite crystal and grains and create other defects. This cumulative structural damage during the reactor lifetime leads to changes in graphite properties, which can alter the ability to support the designed loads. The aim of the present research is to develop a finite-element code using commercially available ABAQUS software for the structural integritymore » analysis of graphite core components under extreme temperature and irradiation conditions. In addition, the Reactor Geometry Generator tool-kit, developed at Argonne National Laboratory, is used to generate finite-element mesh for complex geometries such as fuel bricks with multiple pin holes and coolant flow channels. This paper presents the proposed concept and discusses results of stress analysis simulations of a fuel block with H-451 grade material properties. (authors)« less

Uniform-burning matrix burner

DOEpatents

Bohn, Mark S.; Anselmo, Mark

2001-01-01

Computer simulation was used in the development of an inward-burning, radial matrix gas burner and heat pipe heat exchanger. The burner and exchanger can be used to heat a Stirling engine on cloudy days when a solar dish, the normal source of heat, cannot be used. Geometrical requirements of the application forced the use of the inward burning approach, which presents difficulty in achieving a good flow distribution and air/fuel mixing. The present invention solved the problem by providing a plenum with just the right properties, which include good flow distribution and good air/fuel mixing with minimum residence time. CFD simulations were also used to help design the primary heat exchanger needed for this application which includes a plurality of pins emanating from the heat pipe. The system uses multiple inlet ports, an extended distance from the fuel inlet to the burner matrix, flow divider vanes, and a ring-shaped, porous grid to obtain a high-temperature uniform-heat radial burner. Ideal applications include dish/Stirling engines, steam reforming of hydrocarbons, glass working, and any process requiring high temperature heating of the outside surface of a cylindrical surface.

Influence of the interpellet space to the Instant Release Fraction determination of a commercial UO2 Boiling Water Reactor Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Martínez-Torrents, A.; Serrano-Purroy, D.; Casas, I.; De Pablo, J.

2018-02-01

The contact of the coolant with the fuel pin during irradiation produces a gradient of temperature in the fuel pellet that segregates the radionuclides (RN) depending on its volatility and reactivity. This segregation determines the Instant Release Fraction (IRF), an important source of radiological risk in the performance assessment (PA) of a Deep Geologic Repository (DGR). RN segregation was studied radially in previous papers. In the present work, it was studied axially, taking into special consideration the cutting position of the solid sample to be studied. Iodine and caesium were the RN with the highest release, while the contribution of rubidium, strontium, molybdenum and technetium to the IRF depended on their chemical state. The interpellet presence (known also as dishing) effect was clearly observed for caesium, increasing its release by one order of magnitude. According to these results, one of the major contributions to the IRF comes from the RN trapped in the dishing and has to be considered in the sampling and data interpretation that will be performed for the PA of the DGR.

An experimental study of the distribution of retained xenon in transient-tested UO 2 fuel

NASA Astrophysics Data System (ADS)

Mogensen, M.; Bagger, C.; Walker, C. T.

1993-01-01

XRF and EPMA results for the distribution of retained xenon in twenty fuel pins are surveyed. The aim is to show the progress that has been achieved by combining these methods. One of the main concerns of the paper is the reliability of the XRF and EPMA measurements and the identification, of the principal sources of uncertainty. Another, is the wealth of new mechanistic information that has been acquired by systematically combining XRF and EPMA with quantitative image analysis (QIA) of the local size distribution of the gas bubbles in the fuel. It is shown that by correlating the three data sets it is possible to establish the distribution of retained gas on the grain boundaries and to estimate the pressure of the gas contained in grain boundary bubbles. It is concluded that often gas release during a reactor power transient cannot be predicted on the basis of simple gas diffusion considerations and that it is not possible to derive a gas diffusion coefficent of general relevance from puncturing data.

Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

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

Majed, M.; Morback, G.; Wiman, P.

1995-09-01

The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give largemore » advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.« less

Antimicrobial efficacy of external fixator pins coated with a lipid stabilized hydroxyapatite/chlorhexidine complex to prevent pin tract infection in a goat model.

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

Dejong, E. Schuyler; Deberardino, T. M.; Brooks, D. E.

Background: Pin tract infection is a common complication of external fixation. An antiinfective external fixator pin might help to reduce the incidence of pin tract infection and improve pin fixation. Methods: Stainless steel and titanium external fixator pins, with and without a lipid stabilized hydroxyapatite/chlorhexidine coating, were evaluated in a goat model. Two pins contaminated with an identifiable Staphylococcus aureus strain were inserted into each tibia of 12 goats. The pin sites were examined daily. On day 14, the animals were killed, and the pin tips cultured. Insertion and extraction torques were measured. Results: Infection developed in 100% of uncoatedmore » pins, whereas coated pins demonstrated 4.2% infected, 12.5% colonized, and the remainder, 83.3%, had no growth (p < 0.01). Pin coating decreased the percent loss of fixation torque over uncoated pins (p = 0.04). Conclusion: These results demonstrate that the lipid stabilized hydroxyapatite/chlorhexidine coating was successful in decreasing infection and improving fixation of external fixator pins.« less

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

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

The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, containedmore » in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)« less

Relative sliding durability of two candidate high temperature oxide fiber seal materials

NASA Technical Reports Server (NTRS)

Dellacorte, Christopher; Steinetz, Bruce M.

1991-01-01

A test program to determine the relative sliding durability of two candidate ceramic fibers for high temperature sliding seal applications is described. Pin on disk tests were used to evaluate potential seal materials. Friction during the tests and fiber wear, indicated by the extent of fibers broken in a test bundle or yarn, was measured at the end of a test. In general, friction and wear increase with test temperature. This may be due to a reduction in fiber strength, a change in the surface chemistry at the fiber/counterface interface due to oxidation, adsorption and/or desorption of surface species and, to a lesser extent, an increase in counterface surface roughness due to oxidation at elevated temperatures. The relative fiber durability correlates with tensile strength indicating that tensile data, which is more readily available than sliding durability data, may be useful in predicting fiber wear behavior under various conditions. A simple model developed using dimensional analysis shows that the fiber durability is related to a dimensionless parameter which represents the ratio of the fiber strength to the fiber stresses imposed by sliding.

Sliding durability of candidate seal fiber materials in hydrogen from 25 to 900 C

NASA Technical Reports Server (NTRS)

Dellacorte, Christopher; Steinetz, Bruce M.

1992-01-01

Sliding durability studies of candidate ceramic fibers were conducted in hydrogen to support the high temperature seal development program at NASA LeRC. Pin-on-disk tests were used to measure the friction and durability of a tow or bundle of ceramic fibers in sliding against a superalloy disk. This procedure was used previously to test candidate fibers in an air environment. The fibers based upon mullite (Al2O3-SiO2) chemistry (Nextel 550, 440, and 312) exhibited better durability in hydrogen than in air. HPZ, a complex silicon carboxynitride fiber which showed good durabilty in air, however, showed a significant loss of durability in hot hydrogen. These results are consistent with recent thermodynamic and experimental studies of ceramic compatibility with hydrogen at elevated temperatures. These research results indicate that only oxide fibers display good durability in both air and hydrogen environments. Also, simple, low cost testing in air can provide an adequate data base for initial seal material screening and selection, especially for oxide fiber candidates. The findings of this research provide critical input to the seal design team.

Recognition of white matter bundles using local and global streamline-based registration and clustering.

PubMed

Garyfallidis, Eleftherios; Côté, Marc-Alexandre; Rheault, Francois; Sidhu, Jasmeen; Hau, Janice; Petit, Laurent; Fortin, David; Cunanne, Stephen; Descoteaux, Maxime

2018-04-15

Virtual dissection of diffusion MRI tractograms is cumbersome and needs extensive knowledge of white matter anatomy. This virtual dissection often requires several inclusion and exclusion regions-of-interest that make it a process that is very hard to reproduce across experts. Having automated tools that can extract white matter bundles for tract-based studies of large numbers of people is of great interest for neuroscience and neurosurgical planning. The purpose of our proposed method, named RecoBundles, is to segment white matter bundles and make virtual dissection easier to perform. This can help explore large tractograms from multiple persons directly in their native space. RecoBundles leverages latest state-of-the-art streamline-based registration and clustering to recognize and extract bundles using prior bundle models. RecoBundles uses bundle models as shape priors for detecting similar streamlines and bundles in tractograms. RecoBundles is 100% streamline-based, is efficient to work with millions of streamlines and, most importantly, is robust and adaptive to incomplete data and bundles with missing components. It is also robust to pathological brains with tumors and deformations. We evaluated our results using multiple bundles and showed that RecoBundles is in good agreement with the neuroanatomical experts and generally produced more dense bundles. Across all the different experiments reported in this paper, RecoBundles was able to identify the core parts of the bundles, independently from tractography type (deterministic or probabilistic) or size. Thus, RecoBundles can be a valuable method for exploring tractograms and facilitating tractometry studies. Copyright © 2017 Elsevier Inc. All rights reserved.

On the concept of a filtered bundle

NASA Astrophysics Data System (ADS)

Bruce, Andrew James; Grabowska, Katarzyna; Grabowski, Janusz

We present the notion of a filtered bundle as a generalization of a graded bundle. In particular, we weaken the necessity of the transformation laws for local coordinates to exactly respect the weight of the coordinates by allowing more general polynomial transformation laws. The key examples of such bundles include affine bundles and various jet bundles, both of which play fundamental roles in geometric mechanics and classical field theory. We also present the notion of double filtered bundles which provide natural generalizations of double vector bundles and double affine bundles. Furthermore, we show that the linearization of a filtered bundle — which can be seen as a partial polarization of the admissible changes of local coordinates — is well defined.

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

Ling Zou; Hongbin Zhang; Jess Gehin

A coupled TH/Neutronics/CRUD framework, which is able to simulate the CRUD deposits impact on CIPS phenomenon, was described in this paper. This framework includes the coupling among three essential physics, thermal-hydraulics, CRUD and neutronics. The overall framework was implemented by using the CFD software STAR-CCM+, developing CRUD codes, and using the neutronics code DeCART. The coupling was implemented by exchanging data between softwares using intermediate exchange files. A typical 3 by 3 PWR fuel pin problem was solved under this framework. The problem was solved in a 12 months length period of time. Time-dependent solutions were provided, including CRUD depositsmore » inventory and their distributions on fuels, boron hideout amount inside CRUD deposits, as well as power shape changing over time. The results clearly showed the power shape suppression in regions where CRUD deposits exist, which is a strong indication of CIPS phenomenon.« less

Resonance treatment using pin-based pointwise energy slowing-down method

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

Choi, Sooyoung, E-mail: csy0321@unist.ac.kr; Lee, Changho, E-mail: clee@anl.gov; Lee, Deokjung, E-mail: deokjung@unist.ac.kr

A new resonance self-shielding method using a pointwise energy solution has been developed to overcome the drawbacks of the equivalence theory. The equivalence theory uses a crude resonance scattering source approximation, and assumes a spatially constant scattering source distribution inside a fuel pellet. These two assumptions cause a significant error, in that they overestimate the multi-group effective cross sections, especially for {sup 238}U. The new resonance self-shielding method solves pointwise energy slowing-down equations with a sub-divided fuel rod. The method adopts a shadowing effect correction factor and fictitious moderator material to model a realistic pointwise energy solution. The slowing-down solutionmore » is used to generate the multi-group cross section. With various light water reactor problems, it was demonstrated that the new resonance self-shielding method significantly improved accuracy in the reactor parameter calculation with no compromise in computation time, compared to the equivalence theory.« less

Electric heater for nuclear fuel rod simulators

DOEpatents

McCulloch, Reginald W.; Morgan, Jr., Chester S.; Dial, Ralph E.

1982-01-01

The present invention is directed to an electric cartridge-type heater for use as a simulator for a nuclear fuel pin in reactor studies. The heater comprises an elongated cylindrical housing containing a longitudinally extending helically wound heating element with the heating element radially inwardly separated from the housing. Crushed cold-pressed preforms of boron nitride electrically insulate the heating element from the housing while providing good thermal conductivity. Crushed cold-pressed preforms of magnesia or a magnesia-15 percent boron nitride mixture are disposed in the cavity of the helical heating element. The coefficient of thermal expansion of the magnesia or the magnesia-boron nitride mixture is higher than that of the boron nitride disposed about the heating element for urging the boron nitride radially outwardly against the housing during elevated temperatures to assure adequate thermal contact between the housing and the boron nitride.

Anatomy of the anterior cruciate ligament with regard to its two bundles.

PubMed

Petersen, Wolf; Zantop, Thore

2007-01-01

The anterior cruciate ligament (ACL) consists of two major fiber bundles, namely the anteromedial and posterolateral bundle. When the knee is extended, the posterolateral bundle (PL) is tight and the anteromedial (AM) bundle is moderately lax. As the knee is flexed, the femoral attachment of the ACL becomes a more horizontal orientation; causing the AM bundle to tighten and the PL bundle to relax. There is some degree of variability for the femoral origin of the anterome-dial and posterolateral bundle. The anteromedial bundle is located proximal and anterior in the femoral ACL origin (high and deep in the notch when the knee is flexed at 90 degrees ); the posterolateral bundle starts in the distal and posterior aspect of the femoral ACL origin (shallow and low when the knee is flexed at 90 degrees ). In the frontal plane the anteromedial bundle origin is in the 10:30 clock position and the postero-lateral bundle origin in the 9:30 clock position. At the tibial insertion the ACL fans out to form the foot region. The anteromedial bundle insertion is in the anterior part of the tibial ACL footprint, the posterolateral bundle in the posterior part. While the anteromedial bundle is the primary restraint against anterior tibial translation, the posterolateral bundle tends to stabilize the knee near full extension, particularly against rotatory loads.

Phase equilibria in the UO 2-PuO 2 system under a temperature gradient

NASA Astrophysics Data System (ADS)

Kleykamp, Heiko

2001-04-01

The phase behaviour of U 0.80Pu 0.20O 1.95 was investigated under a steady-state temperature gradient between the solidus and liquidus by a short-time power-to-melt irradiation experiment. The radial U, Pu, Am and O profiles in the fuel pin after redistribution were measured by X-ray microanalysis. During irradiation, an inner fuel melt forms which is separated from the outer solid only by one concentric liquid-solid-phase boundary. The UO 2 concentration increases to 85% and the PuO 2 concentration decreases to 15% on the solid side of the interface. Opposite gradients occur on the liquid side of the interface. The concentration discontinuity is a consequence of the necessary equality of the chemical potentials of UO 2 and PuO 2 on both sides of the phase boundary which corresponds to a 2750°C isotherm. The radial oxygen profile results in an O/(U + Pu) ratio of 2.00 at the fuel surface and 1.92 at the central void of the fuel. The redistribution is caused by the thermal diffusion of oxygen vacancies in the lattice along the temperature gradient. This process is quantified by the heat of transport Q*v which ranges between -10 kJ/mol at the central void and about -230 kJ/mol near the fuel surface.

Mergers, Acquisitions, and Access: STM Publishing Today

NASA Astrophysics Data System (ADS)

Robertson, Kathleen

Electronic publishing is changing the fundamentals of the entire printing/delivery/archive system that has served as the distribution mechanism for scientific research over the last century and a half. The merger-mania of the last 20 years, preprint pools, and publishers' licensing and journals-bundling plans are among the phenomena impacting the scientific information field. Science-Technology-Medical (STM) publishing is experiencing a period of intense consolidation and reorganization. This paper gives an overview of the economic factors fueling these trends, the major STM publishers, and the government regulatory bodies that referee this industry in Europe, Canada, and the USA.

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

NASA Astrophysics Data System (ADS)

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

2017-07-01

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

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

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

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

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

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

DOE PAGES

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

2017-01-05

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

Discovery of novel selenium derivatives as Pin1 inhibitors by high-throughput screening

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

Subedi, Amit; Graduate School of Science and Engineering, Saitama University, Saitama, 338-8570; Shimizu, Takeshi

2016-06-03

Peptidyl prolyl cis/trans isomerization by Pin1 regulates various oncogenic signals during cancer progression, and its inhibition through multiple approaches has established Pin1 as a therapeutic target. However, lack of simplified screening systems has limited the discovery of potent Pin1 inhibitors. We utilized phosphorylation-dependent binding of Pin1 to its specific substrate to develop a screening system for Pin1 inhibitors. Using this system, we screened a chemical library, and identified a novel selenium derivative as Pin1 inhibitor. Based on structure-activity guided chemical synthesis, we developed more potent Pin1 inhibitors that inhibited cancer cell proliferation. -- Highlights: •Novel screening for Pin1 inhibitors basedmore » on Pin1 binding is developed. •A novel selenium compound is discovered as Pin1 inhibitor. •Activity guided chemical synthesis of selenium derivatives resulted potent Pin1 inhibitors.« less

Effect of nitrogen on tensile properties and structures of T-111 (tantalum, 8 percent tungsten, 2 percent hafnium) tubing

NASA Technical Reports Server (NTRS)

Buzzard, R. J.; Metroka, R. R.

1973-01-01

The effect of controlled nitrogen additions was evaluated on the mechanical properties of T-111 (Ta-8W-2Hf) fuel pin cladding material proposed for use in a lithium-cooled nuclear reactor concept. Additions of 80 to 1125 ppm nitrogen resulted in increased strengthening of T-111 tubular section test specimens at temperatures of 25 to 1200 C. Homogeneous distributions of up to 500 ppm nitrogen did not seriously decrease tensile ductility. Both single and two-phase microstructures, with hafnium nitride as the second phase, were evaluated in this study.

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

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

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

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

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

DOE PAGES

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

2016-02-26

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

2011 Annual Criticality Safety Program Performance Summary

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

Andrea Hoffman

The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection,more » an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The TSR limits fuel plate bundles to 1085 grams U-235, which is the maximum loading of an ATR fuel element. The overloaded fuel plate bundle contained 1097 grams U-235 and was assembled under an 1100 gram U-235 limit in 1982. In 2003, the limit was reduced to 1085 grams citing a new criticality safety evaluation for ATR fuel elements. The fuel plate bundle inventories were not checked for compliance prior to implementing the reduced limit. A subsequent review of the NMIS inventory did not identify further violations. Requirements Management - The INL Criticality Safety program is organized and well documented. The source requirements for the INL Criticality Safety Program are from 10 CFR 830.204, DOE Order 420.1B, Chapter III, 'Nuclear Criticality Safety,' ANSI/ANS 8-series Industry Standards, and DOE Standards. These source requirements are documented in LRD-18001, 'INL Criticality Safety Program Requirements Manual.' The majority of the criticality safety source requirements are contained in DOE Order 420.1B because it invokes all of the ANSI/ANS 8-Series Standards. DOE Order 420.1B also invokes several DOE Standards, including DOE-STD-3007, 'Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities.' DOE Order 420.1B contains requirements for DOE 'Heads of Field Elements' to approve the criticality safety program and specific elements of the program, namely, the qualification of criticality staff and the method for preparing criticality safety evaluations. This was accomplished by the approval of SAR-400, 'INL Standardized Nuclear Safety Basis Manual,' Chapter 6, 'Prevention of Inadvertent Criticality.' Chapter 6 of SAR-400 contains sufficient detail and/or reference to the specific DOE and contractor documents that adequately describe the INL Criticality Safety Program per the elements specified in DOE Order 420.1B. The Safety Evaluation Report for SAR-400 specifically recognizes that the approval of SAR-400 approves the INL Criticality Safety Program. No new source requirements were released in 2011. A revision to LRD-18001 is planned for 2012 to clarify design requirements for criticality alarms. Training - Criticality Safety Engineering has developed training and provides training for many employee positions, including fissionable material handlers, facility managers, criticality safety officers, firefighters, and criticality safety engineers. Criticality safety training at the INL is a program strength. A revision to the training module developed in 2010 to supplement MFC certified fissionable material handlers (operators) training was prepared and presented in August of 2011. This training, 'Applied Science of Criticality Safety,' builds upon existing training and gives operators a better understanding of how their criticality controls are derived. Improvements to 00INL189, 'INL Criticality Safety Principles' are planned for 2012 to strengthen fissionable material handler training.« less

CFD modeling of turbulent mixing through vertical pressure tube type boiling water reactor fuel rod bundles with spacer-grids

NASA Astrophysics Data System (ADS)

Verma, Shashi Kant; Sinha, S. L.; Chandraker, D. K.

2018-05-01

Numerical simulation has been carried out for the study of natural mixing of a Tracer (Passive scalar) to describe the development of turbulent diffusion in an injected sub-channel and, afterwards on, cross-mixing between adjacent sub-channels. In this investigation, post benchmark evaluation of the inter-subchannel mixing was initiated to test the ability of state-of-the-art Computational Fluid Dynamics (CFD) codes to numerically predict the important turbulence parameters downstream of a ring type spacer grid in a rod-bundle. A three-dimensional Computational Fluid Dynamics (CFD) tool (STAR-CCM+) was used to model the single phase flow through a 30° segment or 1/12th of the cross segment of a 54-rod bundle with a ring shaped spacer grid. Polyhedrons were used to discretize the computational domain, along with prismatic cells near the walls, with an overall mesh count of 5.2 M cell volumes. The Reynolds Stress Models (RSM) was tested because of RSM accounts for the turbulence anisotropy, to assess their capability in predicting the velocities as well as mass fraction of potassium nitrate measured in the experiment. In this way, the line probes are located in the different position of subchannels which could be used to characterize the progress of the mixing along the flow direction, and the degree of cross-mixing assessed using the quantity of tracer arriving in the neighbouring sub-channels. The predicted dimensionless mixing scalar along the length, however, was in good agreement with the measurements downstream of spacers.

Innovative Tools Advance Revolutionary Weld Technique

NASA Technical Reports Server (NTRS)

2009-01-01

The iconic, orange external tank of the space shuttle launch system not only contains the fuel used by the shuttle s main engines during liftoff but also comprises the shuttle s backbone, supporting the space shuttle orbiter and solid rocket boosters. Given the tank s structural importance and the extreme forces (7.8 million pounds of thrust load) and temperatures it encounters during launch, the welds used to construct the tank must be highly reliable. Variable polarity plasma arc welding, developed for manufacturing the external tank and later employed for building the International Space Station, was until 1994 the best process for joining the aluminum alloys used during construction. That year, Marshall Space Flight Center engineers began experimenting with a relatively new welding technique called friction stir welding (FSW), developed in 1991 by The Welding Institute, of Cambridge, England. FSW differs from traditional fusion welding in that it is a solid-state welding technique, using frictional heat and motion to join structural components without actually melting any of the material. The weld is created by a shouldered pin tool that is plunged into the seam of the materials to be joined. The tool traverses the line while rotating at high speeds, generating friction that heats and softens but does not melt the metal. (The heat produced approaches about 80 percent of the metal s melting temperature.) The pin tool s rotation crushes and stirs the plasticized metal, extruding it along the seam as the tool moves forward. The material cools and consolidates, resulting in a weld with superior mechanical properties as compared to those weld properties of fusion welds. The innovative FSW technology promises a number of attractive benefits. Because the welded materials are not melted, many of the undesirables associated with fusion welding porosity, cracking, shrinkage, and distortion of the weld are minimized or avoided. The process is more energy efficient, safe (no toxic smoke or shielding gas, liquid metal splatter, arcing, dangerous voltage, or radiation), and environmentally sound (no consumables, fumes, or noise) than fusion welding. Under computer control, an automated FSW machine can create welds with high reproducibility, improving efficiency and overall quality of manufactured materials. The process also allows for welding dissimilar metals as well as those metals considered to be "unweldable" such as the 7xxx series aluminum alloys. Its effectiveness and versatility makes FSW useful for aerospace, rail, automotive, marine, and military applications. A downside to FSW, however, is the keyhole opening left in the weld when the FSW pin tool exits the weld joint. This is a significant problem when using the FSW process to join circumferential structures such as pipes and storage containers. Furthermore, weld joints that taper in material thickness also present problems when using the conventional FSW pin tool, because the threaded pin rotating within the weld joint material is a fixed length. There must be capability for the rotating pin to both increase and decrease in length in real time while welding the tapered material. (Both circumferential and tapered thickness weldments are found in the space shuttle external tank.) Marshall engineers addressed both the keyhole and tapered material thickness problems by developing the auto-adjustable pin tool. This unique piece of equipment automatically withdraws the pin into the tool s shoulder for keyhole closeout. In addition, the auto-adjustable pin tool retracts, or shortens, the rotating pin while welding a weld joint that tapers from one thickness to a thinner thickness. This year, the impact of the Marshall innovation was recognized with an "Excellence in Technology Transfer Award" from the Federal Laboratory Consortium.

Results of irradiation of (U0.55Pu0.45)N and (U0.4Pu0.6)N fuels in BOR-60 up to ˜12 at.% burn-up

NASA Astrophysics Data System (ADS)

Rogozkin, B. D.; Stepennova, N. M.; Fedorov, Yu. Ye.; Shishkov, M. G.; Kryukov, F. N.; Kuzmin, S. V.; Nikitin, O. N.; Belyaeva, A. V.; Zabudko, L. M.

2013-09-01

In the article presented are the results of post-irradiation tests of helium bonded fuel pins with mixed mononitride fuel (U0.55Pu0.45)N and (U0.4Pu0.6)N having 85% density irradiated in BOR-60 reactor. Achieved maximum burn-up was, respectively, equal to 9.4 and 12.1 at.% with max linear heat rates 41.9 and 54.5 kW/m. Maximum irradiation dose was 43 dpa. No damage of claddings made of ChS-68 steel (20% cold worked) was observed, and ductility margin existed. Maximum depth of cladding corrosion was within 15 μm. Swelling rates of (U0.4Pu0.6)N and (U0.55Pu0.45)N were, respectively, ˜1.1% and ˜0.68% per 1 at.%. Gas release rate did not exceed 19.3% and 19%. Pattern of porosity distribution in the fuel influenced fuel swelling and gas release rates. Plutonium and uranium are uniformly distributed in the fuel, local minimum values of their content being caused by pores and cracks in the pellets. The observable peaks in content distribution are probably connected with the local formation of isolated phases (e.g. Mo, Pd) while the minimum values refer to fuel pores and cracks. Xenon and cesium tend to migrate from the hot sections of fuel, and therefore their min content is observed in the central section of the fuel pellets. Phase composition of the fuel was determined with X-ray diffractometer. The X-ray patterns of metallographic specimens were obtained by the scanning method (the step was 0.02°, the step exposition was equal to 2 s). From the X-ray diffraction analysis data, it follows that the nitrides of both fuel types have the single-phase structure with an FCC lattice (see Table 6).

Multi level optimization of burnable poison utilization for advanced PWR fuel management

NASA Astrophysics Data System (ADS)

Yilmaz, Serkan

The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as well as minimizing the total Gd amount in the core. The GA code developed many good solutions that satisfy all of the design constraints. For these solutions, the EOC soluble boron concentration changes from 68.9 to 97.2 ppm. It is important to note that the difference of 28.3 ppm between the best and the worst solution in the good solutions region represent the potential of 12.5 Effective-Full-Power-Day (EPFD) savings in cycle length. As a comparison, the best BP loading design has 97.2 ppm soluble boron concentration at EOC while the BP loading with available vendors' U/Gd FA designs has 94.4 ppm SOB at EOC. It was estimated that the difference of 2.8 ppm reflected the potential savings of 1.25 EFPD in cycle length. Moreover, the total Gd amount was reduced by 6.89% in mass that provided extra savings in fuel cost compared to the BP loading pattern with available vendor's U/Gd FA designs. (Abstract shortened by UMI.)

Self-locking double retention redundant pull pin release

NASA Technical Reports Server (NTRS)

Killgrove, Thomas O. (Inventor)

1987-01-01

A double-retention redundant pull pin release system is disclosed. The system responds to a single pull during an intentional release operation. A spiral-threaded main pin is seated in a mating bore in a housing, which main pin has a flange fastened thereon at the part of the main pin which is exterior to the housing. Accidental release tends to rotate the main pin. A secondary pin passes through a slightly oversized opening in the flange and is seated in a second bore in the housing. The pins counteract against one another to prevent accidental release. A frictional lock is shared between the main and secondary pins to enhance further locking of the system. The secondary pin, in response to a first pull, is fully retracted from its bore and flange hole. Thereafter the pull causes the main pin to rotate free of the housing to release, for example, a parachute mechanism.

Establishing the need for an engineering standard for agricultural hitch pins.

PubMed

Deboy, G R; Knapp, W M; Field, W E; Krutz, G W; Corum, C L

2012-04-01

Documented incidents have occurred in which failure or unintentional disengagement of agricultural hitch pins has contributed to property damage and personal injury. An examination of current hitch pin use on a convenience sample of farm operations in Indiana revealed a variety of non-standard, worn and damaged, and inappropriately sized hitch pins in use. Informal interviews with the farm operators confirmed that hitch pin misuse, failure, or disengagement is a relatively widespread problem that remains largely unaddressed. On-site observations also suggested a low use of hitch pin retaining devices or safety chains. A review of prior research revealed that little attention has been given to this problem, and currently no documentation allows for an estimate of the frequency or severity of losses associated with hitch pin misuse, failure, or disengagement. No specific engineering standards were found that directly applied to the design, appropriate selection, or loading capacity of agricultural hitch pins. Major suppliers of replacement hitch pins currently provide little or no information on matching hitch pin size to intended applications, and most replacement hitch pins examined were of foreign origin, with the overwhelming majority imported from China or India. These replacement hitch pins provided no specifications other than diameter, length, and, in some cases, labeling that indicated that the pins had been "heat treated. " Testing of a sample of 11 commercially available replacement hitch pins found variation along the length of the pin shaft and between individual pins in surface hardness, a potential predictor of pin failure. Examination of 17 commercially available replacement pins also revealed a variety of identifiers used to describe pin composition and fabrication methods, e.g., "heat treated." None of the pins examined provided any specifications on loading capacity. It was therefore concluded that there is a need to develop an agricultural hitch pin engineering standard that would reflect current agricultural applications and practices and that would be promoted to both original equipment manufacturers and manufacturers and suppliers of replacement hitch pins. The standard should address the design of composite pins, heat treating, surface hardening, loading capacity and labeling of such, incorporation of unintentional disengagement prevention devices, indicators of the need for replacement due to wear, and safety information that should be included in operator instructions. ASABE is the most appropriate organization to develop such a standard. It was also concluded that agricultural safety and health programs and professionals need to raise the awareness of farmers concerning the appropriate selection and use of agricultural hitch pins, including the need to replace non-standard pins with pins less likely to fail or disengage during use, the need to replace hitch pins with indications of potential failure, and the importance of using appropriate safety chains, especially during transport of equipment behind tractors and trucks on public roads.

Pin-Retraction Mechanism On Quick-Release Cover

NASA Technical Reports Server (NTRS)

Macmartin, Malcolm

1994-01-01

Quick-release cover includes pin-retraction mechanism releasing cover quickly from lower of two sets of pin connections holding cover. Cover released at top by pulling lever as described in "Lever-Arm Pin Puller" (NPO-18788). Removal of cover begins when technician or robot pulls upper-pin-release lever. Cover swings downward until tabs on lower pins are pulled through slots in their receptacles. Lower pins are then free.

Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code

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

Gordienko, P. V., E-mail: gorpavel@vver.kiae.ru; Kotsarev, A. V.; Lizorkin, M. P.

2014-12-15

The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

Electrical Connector Mechanical Seating Sensor

NASA Technical Reports Server (NTRS)

Arens, Ellen; Captain, Janine; Youngquist, Robert

2011-01-01

A sensor provides a measurement of the degree of seating of an electrical connector. This sensor provides a number of discrete distances that a plug is inserted into a socket or receptacle. The number of measurements is equal to the number of pins available in the connector for sensing. On at least two occasions, the Shuttle Program has suffered serious time delays and incurred excessive costs simply because a plug was not seated well within a receptacle. Two methods were designed to address this problem: (1) the resistive pin technique and (2) the discrete length pins technique. In the resistive pin approach, a standard pin in a male connector is replaced with a pin that has a uniform resistivity along its length. This provides a variable resistance on that pin that is dependent on how far the pin is inserted into a socket. This is essentially a linear potentiometer. The discrete approach uses a pin (or a few pins) in the connector as a displacement indicator by truncating the pin length so it sits shorter in the connector than the other pins. A loss of signal on this pin would indicate a discrete amount of displacement of the connector. This approach would only give discrete values of connector displacement, and at least one pin would be needed for each displacement value that would be of interest.

An in Vivo Experimental Comparison of Stainless Steel and Titanium Schanz Screws for External Fixation.

PubMed

Ganser, Antonia; Thompson, Rosemary E; Tami, Ivan; Neuhoff, Dirk; Steiner, Adrian; Ito, Keita

2007-02-01

To compare the clinical benefits of stainless steel (SS) to titanium (Ti) on reducing pin track irritation/infection and pin loosening during external fracture fixation. A tibial gap osteotomy was created in 17 sheep and stabilized with four Schanz screws of either SS or Ti and an external fixation frame. Over the 12 week observation period, pin loosening was assessed by grading the radiolucency around the pins and measuring the extraction torque on pin removal at sacrifice. Irritation/infection was assessed with weekly clinical pin track grading. A histological analysis of the tissue adjacent to the pin site was made to assess biocompatibility. A statistically non-significant trend for less bone resorption around Ti pins was found during the early observation period. However, at sacrifice, there was no difference between the two materials. Also, there was no difference in the extraction torque, and there was similar remodeling and apposition of the bone around the pins. A statistically non-significant trend for more infection about SS pins at sacrifice was found. Histology showed a slightly higher prevalence of reactionary cells in SS samples, but was otherwise not much different than around Ti pins. There is no clinically relevant substantial advantage in using either SS or Ti pins on reducing pin loosening or pin track irritation/infection.

System for Controlling the Stirring Pin of a Friction Stir Welding Apparatus

NASA Technical Reports Server (NTRS)

Ding, R. Jeffrey (Inventor); Romine, Peter L. (Inventor); Oelgoetz, Peter A. (Inventor)

2002-01-01

A control is provided for a friction stir welding apparatus comprising a pin tool which includes a shoulder and a rotating pin extending outwardly from the shoulder of the pin tool and which, in use, is plunged into a workpiece formed contacting workpiece members to stir weld the members together. The control system controls the penetration of the pin tool into the workpiece members which are mounted on a support anvil. The control system includes a pin length controller for controlling pin length relative to the shoulder and for producing a corresponding pin length signal. A pin force sensor senses the force being exerted on the pin during welding and produces a corresponding actual pin force signal. A probe controller controls a probe extending outwardly from the pin, senses a parameter related to the distance between the probe and the supporting anvil and produces a corresponding probe signal. A workpiece standoff sensor senses the standoff distance between the workpiece and the standoff sensor and produces a corresponding standoff signal. A control unit receives the various signals, together with a weld schedule, and, based on these signals and the weld schedule, controls the pin length controller so as to control pin penetration into the workpiece.

Pin routability and pin access analysis on standard cells for layout optimization

NASA Astrophysics Data System (ADS)

Chen, Jian; Wang, Jun; Zhu, ChengYu; Xu, Wei; Li, Shuai; Lin, Eason; Ou, Odie; Lai, Ya-Chieh; Qu, Shengrui

2018-03-01

At advanced process nodes, especially at sub-28nm technology, pin accessibility and routability of standard cells has become one of the most challenging design issues due to the limited router tracks and the increased pin density. If this issue can't be found and resolved during the cell design stage, the pin access problem will be very difficult to be fixed in implementation stage and will make the low efficiency for routing. In this paper, we will introduce a holistic approach for the pin accessibility scoring and routability analysis. For accessibility, the systematic calculator which assigns score for each pin will search the available access points, consider the surrounded router layers, basic design rule and allowed via geometry. Based on the score, the "bad" pins can be found and modified. On pin routability analysis, critical pin points (placing via on this point would lead to failed via insertion) will be searched out for either layout optimization guide or set as OBS for via insertion blocking. By using this pin routability and pin access analysis flow, we are able to improve the library quality and performance.

Flux pinning mechanism in codoped-MgB2 with Al2O3 and SiC

NASA Astrophysics Data System (ADS)

Kiafiroozkoohi, Narjess Sadat; Ghorbani, Shaban Reza; Arabi, Hadi

2018-05-01

MgB2 superconductor samples, co-doped with 0.02 wt of Al2O3 and 0.02 wt SiC, have been examined by M-H loop measurements and calculation of the critical current density based on the Bean model. Normalized volume pinning force, f = F/Fmax, as a function of the reduced magnetic field, h = H/Hmax has been obtained at each temperature. Hughochi flux pinning model, which was included the normal point pinning, the normal surface pinning, and the pinning based on spatial variation in the Ginzburg-Landau parameter, was used to study the flux pinning mechanisms. It was found that the Δκ effect and the normal point pinning mechanisms play the main role in the flux pinning at the magnetic field lower than Hmax and the contribution of the Δκ mechanism increases with the increasing temperature, while the contribution of normal point pinning mechanism decreases. At magnetic field larger than Hmax, the only mechanism that acts as the flux pinning was the normal surface pinning mechanism.

Design of a heatpipe-cooled Mars-surface fission reactor

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.; Reid, Robert S.; Lipinski, Ronald J.; Wright, Steven A.; Talandis, Regina A.

2002-01-01

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars-surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heatpipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed-which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heatpipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heatpipes. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to the Stirling engine. This paper describes the attributes, specifications, and performance of a 15-kWt HOMER reactor. .

Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Bragg-Sitton, Shannon; VanDyke, Melissa

2003-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

NASA Technical Reports Server (NTRS)

Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

2002-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

STEAM STIRRED HOMOGENEOUS NUCLEAR REACTOR

DOEpatents

Busey, H.M.

1958-06-01

A homogeneous nuclear reactor utilizing a selfcirculating liquid fuel is described. The reactor vessel is in the form of a vertically disposed tubular member having the lower end closed by the tube walls and the upper end closed by a removal fianged assembly. A spherical reaction shell is located in the lower end of the vessel and spaced from the inside walls. The reaction shell is perforated on its lower surface and is provided with a bundle of small-diameter tubes extending vertically upward from its top central portion. The reactor vessel is surrounded in the region of the reaction shell by a neutron reflector. The liquid fuel, which may be a solution of enriched uranyl sulfate in ordinary or heavy water, is mainiained at a level within the reactor vessel of approximately the top of the tubes. The heat of the reaction which is created in the critical region within the spherical reaction shell forms steam bubbles which more upwardly through the tubes. The upward movement of these bubbles results in the forcing of the liquid fuel out of the top of these tubes, from where the fuel passes downwardly in the space between the tubes and the vessel wall where it is cooled by heat exchangers. The fuel then re-enters the critical region in the reaction shell through the perforations in the bottom. The upper portion of the reactor vessel is provided with baffles to prevent the liquid fuel from splashing into this region which is also provided with a recombiner apparatus for recombining the radiolytically dissociated moderator vapor and a control means.

A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Gomez, C. F.; Mireles, O. R.; Stewart, E.

2016-01-01

The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

The R&D PERFROI Project on Thermal Mechanical and Thermal Hydraulics Behaviors of a Fuel Rod Assembly during a Loss of Coolant Accident

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

Repetto, G.; Dominguez, C.; Durville, B.

The safety principle in case of a LOCA is to preserve the short and long term coolability of the core. The associated safety requirements are to ensure the resistance of the fuel rods upon quench and post-quench loads and to maintain a coolable geometry in the core. An R&D program has been launched by IRSN with the support of EDF, to perform both experimental and modeling activities in the frame of the LOCA transient, on technical issues such as: - flow blockage within a fuel rods bundle and its potential impact on coolability, - fuel fragment relocation in the balloonedmore » areas: its potential impact on cladding PCT (Peak Cladding Temperature) and on the maximum oxidation rate, - potential loss of cladding integrity upon quench and post-quench loads. The PERFROI project (2014-2019) focusing on the first above issue, is structured in two axes: 1. axis 1: thermal mechanical behavior of deformation and rupture of cladding taking into account the contact between fuel rods; specific research at LaMCoS laboratory focus on the hydrogen behavior in cladding alloys and its impact on the mechanical behavior of the rod; and, 2. axis 2: thermal hydraulics study of a partially blocked region of the core (ballooned area taking into account the fuel relocation with local over power), during cooling phase by water injection; More detailed activities foreseen in collaboration with LEMTA laboratory will focus on the characterization of two phase flows with heat transfer in deformed structures.« less

Signal detection by active, noisy hair bundles

NASA Astrophysics Data System (ADS)

O'Maoiléidigh, Dáibhid; Salvi, Joshua D.; Hudspeth, A. J.

2018-05-01

Vertebrate ears employ hair bundles to transduce mechanical movements into electrical signals, but their performance is limited by noise. Hair bundles are substantially more sensitive to periodic stimulation when they are mechanically active, however, than when they are passive. We developed a model of active hair-bundle mechanics that predicts the conditions under which a bundle is most sensitive to periodic stimulation. The model relies only on the existence of mechanotransduction channels and an active adaptation mechanism that recloses the channels. For a frequency-detuned stimulus, a noisy hair bundle's phase-locked response and degree of entrainment as well as its detection bandwidth are maximized when the bundle exhibits low-amplitude spontaneous oscillations. The phase-locked response and entrainment of a bundle are predicted to peak as functions of the noise level. We confirmed several of these predictions experimentally by periodically forcing hair bundles held near the onset of self-oscillation. A hair bundle's active process amplifies the stimulus preferentially over the noise, allowing the bundle to detect periodic forces less than 1 pN in amplitude. Moreover, the addition of noise can improve a bundle's ability to detect the stimulus. Although, mechanical activity has not yet been observed in mammalian hair bundles, a related model predicts that active but quiescent bundles can oscillate spontaneously when they are loaded by a sufficiently massive object such as the tectorial membrane. Overall, this work indicates that auditory systems rely on active elements, composed of hair cells and their mechanical environment, that operate on the brink of self-oscillation.

ESD testing of the 8S actuator (u)

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

Mchugh, Douglas C

2010-12-03

The 8S actuator is a hot-wire initiated explosive component used to drive the W76-1 2X Acorn 1V valve. It is known to be safe from human electrostatic discharge (ESD) pin-to-pin and all pin-to-cup stimuli as well as 1 amp/1 watt safe. However low impedance (furniture) ESD stimuli applied pin-to-pin has not been evaluated. Components were tested and the results analyzed. The 8S actuator has been shown to be immune to human and severe furniture ESD, whether applied pin-to-pin or pin-to-cup.

Genome-wide identification and evolution of the PIN-FORMED (PIN) gene family in Glycine max.

PubMed

Liu, Yuan; Wei, Haichao

2017-07-01

Soybean (Glycine max) is one of the most important crop plants. Wild and cultivated soybean varieties have significant differences worth further investigation, such as plant morphology, seed size, and seed coat development; these characters may be related to auxin biology. The PIN gene family encodes essential transport proteins in cell-to-cell auxin transport, but little research on soybean PIN genes (GmPIN genes) has been done, especially with respect to the evolution and differences between wild and cultivated soybean. In this study, we retrieved 23 GmPIN genes from the latest updated G. max genome database; six GmPIN protein sequences were changed compared with the previous database. Based on the Plant Genome Duplication Database, 18 GmPIN genes have been involved in segment duplication. Three pairs of GmPIN genes arose after the second soybean genome duplication, and six occurred after the first genome duplication. The duplicated GmPIN genes retained similar expression patterns. All the duplicated GmPIN genes experienced purifying selection (K a /K s < 1) to prevent accumulation of non-synonymous mutations and thus remained more similar. In addition, we also focused on the artificial selection of the soybean PIN genes. Five artificially selected GmPIN genes were identified by comparing the genome sequence of 17 wild and 14 cultivated soybean varieties. Our research provides useful and comprehensive basic information for understanding GmPIN genes.

Reduction of halo pin site morbidity with a new pin care regimen.

PubMed

Kazi, Hussain Anthony; de Matas, Marcus; Pillay, Robin

2013-06-01

A retrospective analysis of halo device associated morbidity over a 4-year period. To assess the impact of a new pin care regimen on halo pin site related morbidity. Halo orthosis treatment still has a role in cervical spine pathology, despite increasing possibilities of open surgical treatment. Published figures for pin site infection range from 12% to 22% with pin loosening from 7% to 50%. We assessed the outcome of a new pin care regimen on morbidity associated with halo spinal orthoses, using a retrospective cohort study from 2001 to 2004. In the last two years, our pin care regimen was changed. This involved pin site care using chlorhexidene & regular torque checking as part of a standard protocol. Previously, povidone iodine was used as skin preparation in theatre, followed by regular sterile saline cleansing when pin sites became encrusted with blood. There were 37 patients in the series, the median age was 49 (range, 22-83) and 20 patients were male. The overall infection rate prior to the new pin care protocol was 30% (n=6) and after the introduction, it dropped to 5.9% (n=1). This difference was statistically significant (p<0.05). Pin loosening occurred in one patient in the group prior to the formal pin care protocol (3%) and none thereafter. Reduced morbidity from halo use can be achieved with a modified pin cleansing and tightening regimen.

Crossed Module Bundle Gerbes; Classification, String Group and Differential Geometry

NASA Astrophysics Data System (ADS)

Jurčo, Branislav

We discuss nonabelian bundle gerbes and their differential geometry using simplicial methods. Associated to any crossed module there is a simplicial group NC, the nerve of the 1-category defined by the crossed module and its geometric realization |NC|. Equivalence classes of principal bundles with structure group |NC| are shown to be one-to-one with stable equivalence classes of what we call crossed module gerbes bundle gerbes. We can also associate to a crossed module a 2-category C'. Then there are two equivalent ways how to view classifying spaces of NC-bundles and hence of |NC|-bundles and crossed module bundle gerbes. We can either apply the W-construction to NC or take the nerve of the 2-category C'. We discuss the string group and string structures from this point of view. Also a simplicial principal bundle can be equipped with a simplicial connection and a B-field. It is shown how in the case of a simplicial principal NC-bundle these simplicial objects give the bundle gerbe connection and the bundle gerbe B-field.

Stand-alone containment analysis of Phébus FPT tests with ASTEC and MELCOR codes: the FPT-2 test.

PubMed

Gonfiotti, Bruno; Paci, Sandro

2018-03-01

During the last 40 years, many studies have been carried out to investigate the different phenomena occurring during a Severe Accident (SA) in a Nuclear Power Plant (NPP). Such efforts have been supported by the execution of different experimental campaigns, and the integral Phébus FP tests were probably some of the most important experiments in this field. In these tests, the degradation of a Pressurized Water Reactor (PWR) fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the findings on these and previous tests, numerical codes such as ASTEC and MELCOR have been developed to analyze the evolution of a SA in real NPPs. After the termination of the Phébus FP campaign, these two codes have been furthermore improved to implement the more recent findings coming from different experimental campaigns. Therefore, continuous verification and validation is still necessary to check that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The aim of the present work is to re-analyze the Phébus FPT-2 test employing the updated ASTEC and MELCOR code versions. The analysis focuses on the stand-alone containment aspects of this test, and three different spatial nodalizations of the containment vessel (CV) have been developed. The paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products (FP) behavior. When possible, a comparison among the results obtained during this work and by different authors in previous work is also performed. This paper is part of a series of publications covering the four Phébus FP tests using a PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3, excluding the FPT-4 one, related to the study of the release of low-volatility FP and transuranic elements from a debris bed and a pool of melted fuel.

Anatomical approach to permanent His bundle pacing: Optimizing His bundle capture.

PubMed

Vijayaraman, Pugazhendhi; Dandamudi, Gopi

2016-01-01

Permanent His bundle pacing is a physiological alternative to right ventricular pacing. In this article we describe our approach to His bundle pacing in patients with AV nodal and intra-Hisian conduction disease. It is essential for the implanters to understand the anatomic variations of the His bundle course and its effect on the type of His bundle pacing achieved. We describe several case examples to illustrate our anatomical approach to permanent His bundle pacing in this article. Copyright © 2016 Elsevier Inc. All rights reserved.

Analytical methods in the high conversion reactor core design

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

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

The prolyl isomerase Pin1 increases β-cell proliferation and enhances insulin secretion.

PubMed

Nakatsu, Yusuke; Mori, Keiichi; Matsunaga, Yasuka; Yamamotoya, Takeshi; Ueda, Koji; Inoue, Yuki; Mitsuzaki-Miyoshi, Keiko; Sakoda, Hideyuki; Fujishiro, Midori; Yamaguchi, Suguru; Kushiyama, Akifumi; Ono, Hiraku; Ishihara, Hisamitsu; Asano, Tomoichiro

2017-07-14

The prolyl isomerase Pin1 binds to the phosphorylated Ser/Thr-Pro motif of target proteins and enhances their cis-trans conversion. This report is the first to show that Pin1 expression in pancreatic β cells is markedly elevated by high-fat diet feeding and in ob/ob mice. To elucidate the role of Pin1 in pancreatic β cells, we generated β-cell-specific Pin1 KO (βPin1 KO) mice. These mutant mice showed exacerbation of glucose intolerance but had normal insulin sensitivity. We identified two independent factors underlying impaired insulin secretion in the βPin1 KO mice. Pin1 enhanced pancreatic β-cell proliferation, as indicated by a reduced β-cell mass in βPin1 KO mice compared with control mice. Moreover, a diet high in fat and sucrose failed to increase pancreatic β-cell growth in the βPin1 KO mice, an observation to which up-regulation of the cell cycle protein cyclin D appeared to contribute. The other role of Pin1 was to activate the insulin-secretory step: Pin1 KO β cells showed impairments in glucose- and KCl-induced elevation of the intracellular Ca 2+ concentration and insulin secretion. We also identified salt-inducible kinase 2 (SIK2) as a Pin1-binding protein that affected the regulation of Ca 2+ influx and found Pin1 to enhance SIK2 kinase activity, resulting in a decrease in p35 protein, a negative regulator of Ca 2+ influx. Taken together, our observations demonstrate critical roles of Pin1 in pancreatic β cells and that Pin1 both promotes β-cell proliferation and activates insulin secretion. © 2017 by The American Society for Biochemistry and Molecular Biology, Inc.

PWR and BWR spent fuel assembly gamma spectra measurements

NASA Astrophysics Data System (ADS)

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

2016-10-01

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

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

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

Katoh, Yutai; Wilson, Dane F; Forsberg, Charles W

2007-09-01

The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) compositesmore » are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.« less

PWR and BWR spent fuel assembly gamma spectra measurements

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

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

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

Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

NASA Astrophysics Data System (ADS)

Tobin, S. J.; Menlove, H. O.; Swinhoe, M. T.; Schear, M. A.

2011-10-01

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is "technology driven" in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

PWR and BWR spent fuel assembly gamma spectra measurements

DOE PAGES

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

2016-07-17

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

Target assembly

DOEpatents

Lewis, Richard A.

1980-01-01

A target for a proton beam which is capable of generating neutrons for absorption in a breeding blanket includes a plurality of solid pins formed of a neutron emissive target material disposed parallel to the path of the beam and which are arranged axially in a plurality of layers so that pins in each layer are offset with respect to pins in all other layers, enough layers being used so that each proton in the beam will strike at least one pin with means being provided to cool the pins. For a 300 mA, 1 GeV beam (300 MW), stainless steel pins, 12 inches long and 0.23 inches in diameter are arranged in triangular array in six layers with one sixth of the pins in each layer, the number of pins being such that the entire cross sectional area of the beam is covered by the pins with minimum overlap of pins.

The effect of temperature on pinning mechanisms in HTS composites

NASA Astrophysics Data System (ADS)

Sotnikova, A. P.; Rudnev, I. A.

2016-09-01

Pinning mechanism in samples of second generation tapes (2G) of high-temperature superconductors (HTS) was studied The critical current and the pinning force were calculated from the magnetization curves measured in the temperature range of 4.2 - 77 K in magnetic fields up to 14 Tesla using vibration sample magnetometer. To determine the pinning mechanism the dependences of pinning force on magnetic field were constructed according to the Dew-Hughes model and Kramer's rule. The obtained dependences revealed a significant influence of the temperature on effectiveness of different types of pinning. At low temperatures the 2G HTS tapes of different manufacturers demonstrated an equal efficiency of the pinning centers but with temperature increase the differences in pinning mechanisms as well as in properties and effectiveness of the pinning centers become obvious. The influence of the pinning mechanism on the energy losses in HTS tapes was shown.

Pin fin compliant heat sink with enhanced flexibility

DOEpatents

Schultz, Mark D.

2018-04-10

Heat sinks and methods of using the same include a top and bottom plate, at least one of which has a plurality of pin contacts flexibly connected to one another, where the plurality of pin contacts have vertical and lateral flexibility with respect to one another; and pin slice layers, each having multiple pin slices, arranged vertically between the top and bottom plates such that the plurality of pin slices form substantially vertical pins connecting the top and bottom plates.

Warps, grids and curvature in triple vector bundles

NASA Astrophysics Data System (ADS)

Flari, Magdalini K.; Mackenzie, Kirill

2018-06-01

A triple vector bundle is a cube of vector bundle structures which commute in the (strict) categorical sense. A grid in a triple vector bundle is a collection of sections of each bundle structure with certain linearity properties. A grid provides two routes around each face of the triple vector bundle, and six routes from the base manifold to the total manifold; the warps measure the lack of commutativity of these routes. In this paper we first prove that the sum of the warps in a triple vector bundle is zero. The proof we give is intrinsic and, we believe, clearer than the proof using decompositions given earlier by one of us. We apply this result to the triple tangent bundle T^3M of a manifold and deduce (as earlier) the Jacobi identity. We further apply the result to the triple vector bundle T^2A for a vector bundle A using a connection in A to define a grid in T^2A . In this case the curvature emerges from the warp theorem.

Cytokinins influence root gravitropism via differential regulation of auxin transporter expression and localization in Arabidopsis.

PubMed

Pernisova, Marketa; Prat, Tomas; Grones, Peter; Harustiakova, Danka; Matonohova, Martina; Spichal, Lukas; Nodzynski, Tomasz; Friml, Jiri; Hejatko, Jan

2016-10-01

Redirection of intercellular auxin fluxes via relocalization of the PIN-FORMED 3 (PIN3) and PIN7 auxin efflux carriers has been suggested to be necessary for the root gravitropic response. Cytokinins have also been proposed to play a role in controlling root gravitropism, but conclusive evidence is lacking. We present a detailed study of the dynamics of root bending early after gravistimulation, which revealed a delayed gravitropic response in transgenic lines with depleted endogenous cytokinins (Pro35S:AtCKX) and cytokinin signaling mutants. Pro35S:AtCKX lines, as well as a cytokinin receptor mutant ahk3, showed aberrations in the auxin response distribution in columella cells consistent with defects in the auxin transport machinery. Using in vivo real-time imaging of PIN3-GFP and PIN7-GFP in AtCKX3 overexpression and ahk3 backgrounds, we observed wild-type-like relocalization of PIN proteins in the columella early after gravistimulation, with gravity-induced relocalization of PIN7 faster than that of PIN3. Nonetheless, the cellular distribution of PIN3 and PIN7 and expression of PIN7 and the auxin influx carrier AUX1 was affected in AtCKX overexpression lines. Based on the retained cytokinin sensitivity in pin3 pin4 pin7 mutant, we propose the AUX1-mediated auxin transport rather than columella-located PIN proteins as a target of endogenous cytokinins in the control of root gravitropism. © 2016 The Authors. New Phytologist © 2016 New Phytologist Trust.

Analysis of key safety metrics of thorium utilization in LWRs

DOE PAGES

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

2016-04-08

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

Optimization of intrinsic layer thickness, dopant layer thickness and concentration for a-SiC/a-SiGe multilayer solar cell efficiency performance using Silvaco software

NASA Astrophysics Data System (ADS)

Yuan, Wong Wei; Natashah Norizan, Mohd; Salwani Mohamad, Ili; Jamalullail, Nurnaeimah; Hidayah Saad, Nor

2017-11-01

Solar cell is expanding as green renewable alternative to conventional fossil fuel electricity generation, but compared to other land-used electrical generators, it is a comparative beginner. Many applications covered by solar cells starting from low power mobile devices, terrestrial, satellites and many more. To date, the highest efficiency solar cell is given by GaAs based multilayer solar cell. However, this material is very expensive in fabrication and material costs compared to silicon which is cheaper due to the abundance of supply. Thus, this research is devoted to develop multilayer solar cell by combining two different layers of P-I-N structures with silicon carbide and silicon germanium. This research focused on optimising the intrinsic layer thickness, p-doped layer thickness and concentration, n-doped layer thickness and concentration in achieving the highest efficiency. As a result, both single layer a-SiC and a-SiGe showed positive efficiency improvement with the record of 27.19% and 9.07% respectively via parametric optimization. The optimized parameters is then applied on both SiC and SiGe P-I-N layers and resulted the convincing efficiency of 33.80%.

KSC-08pd0006

NASA Image and Video Library

2008-01-04

KENNEDY SPACE CENTER, FLA. -- At a lab at NASA's Kennedy Space Center, Bob Arp, an aerospace technician with the United Launch Alliance, solders a pin to the socket of the replacement feed-through connector that will be installed in the external fuel tank for space shuttle Atlantis' STS-122 mission. The technician performed this exacting task on the Centaur upper stage for Atlas and Titan launches in 1994 and was specifically chosen for the task. Soldering the connector pins and sockets together addresses the most likely cause of a problem in the engine cutoff sensor system, or ECO system. Some of the tank's ECO sensors failed during propellant tanking for launch attempts on Dec. 6 and Dec. 9. Results of a tanking test on Dec. 18 pointed to an open circuit in the feed-through connector wiring, which is located at the base of the tank. The feed-through connector passes the wires from the inside of the tank to the outside. After the soldering is completed and the connector is reinstalled, shuttle program managers will decide on how to proceed. The launch date for mission STS-122 is under review. Photo credit: NASA/Kim Shiflett

KSC-08pd0004

NASA Image and Video Library

2008-01-04

KENNEDY SPACE CENTER, FLA. -- At a lab at NASA's Kennedy Space Center, aerospace technicians with the United Launch Alliance inspect an electrical wiring harness that has been inserted into a replacement feed-through connector during preparations to solder the pins to the socket of the connector that will be installed in the external fuel tank for space shuttle Atlantis' STS-122 mission. The technicians performed this exacting task on the Centaur upper stage for Atlas and Titan launches in 1994 and were specifically chosen for the task. Soldering the connector pins and sockets together addresses the most likely cause of a problem in the engine cutoff sensor system, or ECO system. Some of the tank's ECO sensors failed during propellant tanking for launch attempts on Dec. 6 and Dec. 9. Results of a tanking test on Dec. 18 pointed to an open circuit in the feed-through connector wiring, which is located at the base of the tank. The feed-through connector passes the wires from the inside of the tank to the outside. After the soldering is completed and the connector is reinstalled, shuttle program managers will decide on how to proceed. The launch date for mission STS-122 is under review. Photo credit: NASA/Kim Shiflett

KSC-08pd0001

NASA Image and Video Library

2008-01-04

KENNEDY SPACE CENTER, FLA. -- At a lab at NASA's Kennedy Space Center, a Lockheed Martin technician prepares an electrical wiring harness during a procedure to solder the pins to the socket of the replacement feed-through connector that will be installed in the external fuel tank for space shuttle Atlantis' STS-122 mission. Two United Launch Alliance technicians, who performed this exacting task on the Centaur upper stage for Atlas and Titan launches in 1994, will be doing the soldering. Soldering the connector pins and sockets together addresses the most likely cause of a problem in the engine cutoff sensor system, or ECO system. Some of the tank's ECO sensors failed during propellant tanking for launch attempts on Dec. 6 and Dec. 9. Results of a tanking test on Dec. 18 pointed to an open circuit in the feed-through connector wiring, which is located at the base of the tank. The feed-through connector passes the wires from the inside of the tank to the outside. After the soldering is completed and the connector is reinstalled, shuttle program managers will decide on how to proceed. The launch date for mission STS-122 is under review. Photo credit: NASA/Kim Shiflett

KSC-08pd0009

NASA Image and Video Library

2008-01-04

KENNEDY SPACE CENTER, FLA. -- At a lab at NASA's Kennedy Space Center, Bob Arp, an aerospace technician with the United Launch Alliance, examinies the pins remaining to be soldered to the socket of the replacement feed-through connector that will be installed in the external fuel tank for space shuttle Atlantis' STS-122 mission. The technician performed this exacting task on the Centaur upper stage for Atlas and Titan launches in 1994 and was specifically chosen for the task. Soldering the connector pins and sockets together addresses the most likely cause of a problem in the engine cutoff sensor system, or ECO system. Some of the tank's ECO sensors failed during propellant tanking for launch attempts on Dec. 6 and Dec. 9. Results of a tanking test on Dec. 18 pointed to an open circuit in the feed-through connector wiring, which is located at the base of the tank. The feed-through connector passes the wires from the inside of the tank to the outside. After the soldering is completed and the connector is reinstalled, shuttle program managers will decide on how to proceed. The launch date for mission STS-122 is under review. Photo credit: NASA/Kim Shiflett

KSC-08pd0003

NASA Image and Video Library

2008-01-04

KENNEDY SPACE CENTER, FLA. -- At a lab at NASA's Kennedy Space Center, Kevin Wyckoff, an aerospace technician with the United Launch Alliance, inserts an electrical wiring harness into a replacement feed-through connector during preparations to solder the pins to the socket of the connector. The connector will later be installed in the external fuel tank for space shuttle Atlantis' STS-122 mission. The technician performed this exacting task on the Centaur upper stage for Atlas and Titan launches in 1994 and was specifically chosen for the task. Soldering the connector pins and sockets together addresses the most likely cause of a problem in the engine cutoff sensor system, or ECO system. Some of the tank's ECO sensors failed during propellant tanking for launch attempts on Dec. 6 and Dec. 9. Results of a tanking test on Dec. 18 pointed to an open circuit in the feed-through connector wiring, which is located at the base of the tank. The feed-through connector passes the wires from the inside of the tank to the outside. After the soldering is completed and the connector is reinstalled, shuttle program managers will decide on how to proceed. The launch date for mission STS-122 is under review. Photo credit: NASA/Kim Shiflett

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Steady-state stiffness of utricular hair cells depends on macular location and hair bundle structure.

PubMed

Spoon, Corrie; Moravec, W J; Rowe, M H; Grant, J W; Peterson, E H

2011-12-01

Spatial and temporal properties of head movement are encoded by vestibular hair cells in the inner ear. One of the most striking features of these receptors is the orderly structural variation in their mechanoreceptive hair bundles, but the functional significance of this diversity is poorly understood. We tested the hypothesis that hair bundle structure is a significant contributor to hair bundle mechanics by comparing structure and steady-state stiffness of 73 hair bundles at varying locations on the utricular macula. Our first major finding is that stiffness of utricular hair bundles varies systematically with macular locus. Stiffness values are highest in the striola, near the line of hair bundle polarity reversal, and decline exponentially toward the medial extrastriola. Striolar bundles are significantly more stiff than those in medial (median: 8.9 μN/m) and lateral (2.0 μN/m) extrastriolae. Within the striola, bundle stiffness is greatest in zone 2 (106.4 μN/m), a band of type II hair cells, and significantly less in zone 3 (30.6 μN/m), which contains the only type I hair cells in the macula. Bathing bundles in media that break interciliary links produced changes in bundle stiffness with predictable time course and magnitude, suggesting that links were intact in our standard media and contributed normally to bundle stiffness during measurements. Our second major finding is that bundle structure is a significant predictor of steady-state stiffness: the heights of kinocilia and the tallest stereocilia are the most important determinants of bundle stiffness. Our results suggest 1) a functional interpretation of bundle height variability in vertebrate vestibular organs, 2) a role for the striola in detecting onset of head movement, and 3) the hypothesis that differences in bundle stiffness contribute to diversity in afferent response dynamics.

Steady-state stiffness of utricular hair cells depends on macular location and hair bundle structure

PubMed Central

Spoon, Corrie; Moravec, W. J.; Rowe, M. H.; Grant, J. W.

2011-01-01

Spatial and temporal properties of head movement are encoded by vestibular hair cells in the inner ear. One of the most striking features of these receptors is the orderly structural variation in their mechanoreceptive hair bundles, but the functional significance of this diversity is poorly understood. We tested the hypothesis that hair bundle structure is a significant contributor to hair bundle mechanics by comparing structure and steady-state stiffness of 73 hair bundles at varying locations on the utricular macula. Our first major finding is that stiffness of utricular hair bundles varies systematically with macular locus. Stiffness values are highest in the striola, near the line of hair bundle polarity reversal, and decline exponentially toward the medial extrastriola. Striolar bundles are significantly more stiff than those in medial (median: 8.9 μN/m) and lateral (2.0 μN/m) extrastriolae. Within the striola, bundle stiffness is greatest in zone 2 (106.4 μN/m), a band of type II hair cells, and significantly less in zone 3 (30.6 μN/m), which contains the only type I hair cells in the macula. Bathing bundles in media that break interciliary links produced changes in bundle stiffness with predictable time course and magnitude, suggesting that links were intact in our standard media and contributed normally to bundle stiffness during measurements. Our second major finding is that bundle structure is a significant predictor of steady-state stiffness: the heights of kinocilia and the tallest stereocilia are the most important determinants of bundle stiffness. Our results suggest 1) a functional interpretation of bundle height variability in vertebrate vestibular organs, 2) a role for the striola in detecting onset of head movement, and 3) the hypothesis that differences in bundle stiffness contribute to diversity in afferent response dynamics. PMID:21918003

Double-bundle anterior cruciate ligament reconstruction is superior to single-bundle reconstruction in terms of revision frequency: a study of 22,460 patients from the Swedish National Knee Ligament Register.

PubMed

Svantesson, Eleonor; Sundemo, David; Hamrin Senorski, Eric; Alentorn-Geli, Eduard; Musahl, Volker; Fu, Freddie H; Desai, Neel; Stålman, Anders; Samuelsson, Kristian

2017-12-01

Studies comparing single- and double-bundle anterior cruciate ligament (ACL) reconstructions often include a combined analysis of anatomic and non-anatomic techniques. The purpose of this study was to compare the revision rates between single- and double-bundle ACL reconstructions in the Swedish National Knee Ligament Register with regard to surgical variables as determined by the anatomic ACL reconstruction scoring checklist (AARSC). Patients from the Swedish National Knee Ligament Register who underwent either single- or double-bundle ACL reconstruction with hamstring tendon autograft during the period 2007-2014 were included. The follow-up period started with primary ACL reconstruction, and the outcome measure was set as revision surgery. An online questionnaire based on the items of the AARSC was used to determine the surgical technique implemented in the single-bundle procedures. These were organized into subgroups based on surgical variables, and the revision rates were compared with the double-bundle ACL reconstruction. Hazard ratios (HR) with 95% confidence interval (CI) was calculated and adjusted for confounders by Cox regression. A total of 22,460 patients were included in the study, of which 21,846 were single-bundle and 614 were double-bundle ACL reconstruction. Double-bundle ACL reconstruction had a revision frequency of 2.0% (n = 12) and single-bundle 3.2% (n = 689). Single-bundle reconstruction had an increased risk of revision surgery compared with double-bundle [adjusted HR 1.98 (95% CI 1.12-3.51), p = 0.019]. The subgroup analysis showed a significantly increased risk of revision surgery in patients undergoing single-bundle with anatomic technique using transportal drilling [adjusted HR 2.51 (95% CI 1.39-4.54), p = 0.002] compared with double-bundle ACL reconstruction. Utilizing a more complete anatomic technique according to the AARSC lowered the hazard rate considerably when transportal drilling was performed but still resulted in significantly increased risk of revision surgery compared with double-bundle ACL reconstruction [adjusted HR 1.87 (95% CI 1.04-3.38), p = 0.037]. Double-bundle ACL reconstruction is associated with a lower risk of revision surgery than single-bundle ACL reconstruction. Single-bundle procedures performed using transportal femoral drilling technique had significantly higher risk of revision surgery compared with double-bundle. However, a reference reconstruction with transportal drilling defined as a more complete anatomic reconstruction reduces the risk of revision surgery considerably. III.

The Effect of Insertion Technique on Temperatures for Standard and Self-Drilling External Fixation Pins.

PubMed

Manoogian, Sarah; Lee, Adam K; Widmaier, James C

2017-08-01

No studies have assessed the effects of parameters associated with insertion temperature in modern self-drilling external fixation pins. The current study assessed how varying the presence of irrigation, insertion speed, and force impacted the insertion temperatures of 2 types of standard and self-drilling external fixation half pins. Seventy tests were conducted with 10 trials for 4 conditions on self-drilling pins, and 3 conditions for standard pins. Each test used a thermocouple inside the pin to measure temperature rise during insertion. Adding irrigation to the standard pin insertion significantly lowered the maximum temperature (P <0.001). Lowering the applied force for the standard pin did not have a significant change in temperature rise. Applying irrigation during the self-drilling pin tests dropped average rise in temperature from 151.3 ± 21.6°C to 124.1 ± 15.3°C (P = 0.005). When the self-drilling pin insertion was decreased considerably from 360 to 60 rpm, the temperature decreased significantly from 151.3 ± 21.6°C to 109.6 ± 14.0°C (P <0.001). When the force applied increased significantly, the corresponding self-drilling pin temperature increase was not significant. The standard pin had lower peak temperatures than the self-drilling pin for all conditions. Moreover, slowing down the insertion speed and adding irrigation helped mitigate the temperature increase of both pin types during insertion.

PinX1 Is a Potential Prognostic Factor for Non-Small-Cell Lung Cancer and Inhibits Cell Proliferation and Migration

PubMed Central

Wang, Shengguang; Zhang, Hua; Zhu, Jianquan; Li, Chenguang; Zhu, Jinfang; Shi, Bowen; Zhang, Bin

2017-01-01

PinX1 has been identified as a suppressor of telomerase enzymatic activity. However, the tumour-suppressive roles of PinX1 in different types of human cancers are unclear. PinX1 expression status and its correlation with clinicopathological features in non-small-cell lung cancer (NSCLC) have not been investigated. Accordingly, in this study, we aimed to evaluate the roles of PinX1 in NSCLC. PinX1 expression status was examined by immunohistochemistry using tissue microarray from a total of 158 patients. Correlations among PinX1 expression, clinicopathological variables, and patient survival were analysed. Furthermore, we overexpressed PinX1 in NSCLC cells and tested telomerase activity using real-time quantitative telomeric repeat amplification protocol (qTRAP) assays. Proliferation and migration of NSCLC cells were examined using the MTS method, wound healing assays, and transwell assays, respectively. Our results showed that negative PinX1 expression was associated with a poor prognosis in NSCLC. Sex, smoking status, lymph gland status, subcarinal lymph node status, pathological stage, and PinX1 expression were related to survival. PinX1 was not an independent prognostic factor in NSCLC. PinX1 overexpression inhibited proliferation and migration in NSCLC cells by suppressing telomerase activity. Our findings suggested that PinX1 could be a potential tumour suppressor in NSCLC and that loss of PinX1 promoted NSCLC progression. PMID:28815183

Effects of various tool pin profiles on mechanical and metallurgical properties of friction stir welded joints of cryorolled AA2219 aluminium alloy

NASA Astrophysics Data System (ADS)

Kamal Babu, Karupannan; Panneerselvam, Kavan; Sathiya, Paulraj; Noorul Haq, Abdul Haq; Sundarrajan, Srinivasan; Mastanaiah, Potta; Srinivasa Murthy, Chunduri Venkata

2018-02-01

Friction stir welding (FSW) process was conducted on cryorolled (CR) AA2219 plate using different tool pin profiles such as cylindrical pin, threaded cylindrical pin, square pin and hexagonal pin profiles. The FSW was carried out with pairs of 6 mm thick CR aluminium plates with different tool pin profiles. The different tool pin profile weld portions' behaviors like mechanical (tensile strength, impact and hardness) and metallurgical characteristics were analyzed. The results of the mechanical analysis revealed that the joint made by the hexagonal pin tool had good strength compared to other pin profiles. This was due to the pulsating action and material flow of the tool resulting in dynamic recrystallization in the weld zone. This was confirmed by the ultra fine grain structure formation in Weld Nugget (WN) of hexagonal pin tool joint with a higher percentage of precipitate dissolution. The fractograph of the hexagonal tool pin weld portion confirmed the finer dimple structure morphology without having any interior defect compared to other tool pin profiles. The lowest weld joint strength was obtained from cylindrical pin profile weld joint due to insufficient material flow during welding. The Transmission Electron Microscope and EDX analysis showed the dissolution of the metastable θ″, θ' (Al2Cu) partial precipitates in the WN and proved the influence of metastable precipitates on enhancement of mechanical behavior of weld. The XRD results also confirmed the Al2Cu precipitation dissolution in the weld zone.

A Comprehensive Reanalysis of the Distal Iliotibial Band: Quantitative Anatomy, Radiographic Markers, and Biomechanical Properties.

PubMed

Godin, Jonathan A; Chahla, Jorge; Moatshe, Gilbert; Kruckeberg, Bradley M; Muckenhirn, Kyle J; Vap, Alexander R; Geeslin, Andrew G; LaPrade, Robert F

2017-09-01

The qualitative anatomy of the distal iliotibial band (ITB) has previously been described. However, a comprehensive characterization of the quantitative anatomic, radiographic, and biomechanical properties of the Kaplan fibers of the deep distal ITB has not yet been established. It is paramount to delineate these characteristics to fully understand the distal ITB's contribution to rotational knee stability. Purpose/Hypothesis: There were 2 distinct purposes for this study: (1) to perform a quantitative anatomic and radiographic evaluation of the distal ITB's attachment sites and their relationships to pertinent osseous and soft tissue landmarks, and (2) to quantify the biomechanical properties of the deep (Kaplan) fibers of the distal ITB. It was hypothesized that the distal ITB has definable parameters concerning its anatomic attachments and consistent relationships to surgically pertinent landmarks with correlating plain radiographic findings. In addition, it was hypothesized that the biomechanical properties of the Kaplan fibers would support their role as important restraints against internal rotation. Descriptive laboratory study. Ten nonpaired, fresh-frozen human cadaveric knees (mean age, 61.1 years; range, 54-65 years) were dissected for anatomic and radiographic purposes. A coordinate measuring device quantified the attachment areas of the distal ITB to the distal femur, patella, and proximal tibia and their relationships to pertinent bony landmarks. A radiographic analysis was performed by inserting pins into the attachment sites of relevant anatomic structures to assess their location relative to pertinent bony landmarks with fluoroscopic guidance. A further biomechanical assessment of 10 cadaveric knees quantified the load to failure and stiffness of the Kaplan fibers' insertion on the distal femur after a preconditioning protocol. Two separate deep (Kaplan) fiber bundles were identified with attachments to 2 newly identified femoral bony prominences (ridges). The proximal and distal bundles inserted on the distal femur 53.6 mm (95% CI, 50.7-56.6 mm) and 31.4 mm (95% CI, 27.3-35.5 mm) proximal to the lateral epicondyle, respectively. The centers of the bundle insertions were 22.5 mm (95% CI, 19.1-25.9 mm) apart. The total insertion area of the distal ITB on the proximal tibia was 429.1 mm 2 (95% CI, 349.2-509.1 mm 2 ). A distinct capsulo-osseous layer of the distal ITB was also identified that was intimately related to the lateral knee capsule. Its origin was in close proximity to the lateral gastrocnemius tubercle, and it inserted on the proximal tibia at the lateral tibial tubercle between the fibular head and the Gerdy tubercle. Radiographic analysis supported the quantitative anatomic findings. The mean maximum load during pull-to-failure testing was 71.3 N (95% CI, 41.2-101.4 N) and 170.2 N (95% CI, 123.6-216.8 N) for the proximal and distal Kaplan bundles, respectively. The most important finding of this study was that 2 distinct deep bundles (Kaplan fibers) of the distal ITB were identified. Each bundle of the deep layer of the ITB was associated with a newly identified distinct bony ridge. Radiographic analysis confirmed the measurements previously recorded and established reproducible landmarks for the newly described structures. Biomechanical testing revealed that the Kaplan fibers had a strong attachment to the distal femur, thereby supporting a role in rotational knee stability. The identification of 2 distinct deep fiber (Kaplan) attachments clarifies the function of the ITB more definitively. The results also support the role of the ITB in rotatory knee stability because of the fibers' vectors and their identified maximum loads. These findings provide the anatomic and biomechanical foundation needed for the development of reconstruction or repair techniques to anatomically address these deficiencies in knee ligament injuries.

Moisture separator reheater with round tube bundle

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

Byerley, W. M.

1984-11-27

A moisture separator reheater having a central chamber with cylindrical wall protions and a generally round tube bundle, the tube bundle having arcuate plates disposed on each side of the bundle which form a wrapper on each side of the bundle and having a tongue and groove juncture between the wrapper and cylindrical wall portions to provide a seal therebetween and a track for installing and removing the tube bundle from the central chamber.

["Habitual" left branch block alternating with 2 "disguised" bracnch block].

PubMed

Lévy, S; Jullien, G; Mathieu, P; Mostefa, S; Gérard, R

1976-10-01

Two cases of alternating left bundle branch block and "masquerading block" (with left bundle branch morphology in the stnadard leads and right bundle branch block morphology in the precordial leads) were studied by serial tracings and his bundle electrocardiography. In case 1 "the masquerading" block was associated with a first degree AV block related to a prolongation of HV interval. This case is to our knowledge the first cas of alternating bundle branch block in which his bundle activity was recorded in man. In case 2, the patient had atrial fibrilation and His bundle recordings were performed while differents degrees of left bundle branch block were present: The mechanism of the alternation and the concept of "masquerading" block are discussed. It is suggested that this type of block represents a right bundle branch block associated with severe lesions of the "left system".

Characterization of active hair-bundle motility by a mechanical-load clamp

NASA Astrophysics Data System (ADS)

Salvi, Joshua D.; Maoiléidigh, Dáibhid Ó.; Fabella, Brian A.; Tobin, Mélanie; Hudspeth, A. J.

2015-12-01

Active hair-bundle motility endows hair cells with several traits that augment auditory stimuli. The activity of a hair bundle might be controlled by adjusting its mechanical properties. Indeed, the mechanical properties of bundles vary between different organisms and along the tonotopic axis of a single auditory organ. Motivated by these biological differences and a dynamical model of hair-bundle motility, we explore how adjusting the mass, drag, stiffness, and offset force applied to a bundle control its dynamics and response to external perturbations. Utilizing a mechanical-load clamp, we systematically mapped the two-dimensional state diagram of a hair bundle. The clamp system used a real-time processor to tightly control each of the virtual mechanical elements. Increasing the stiffness of a hair bundle advances its operating point from a spontaneously oscillating regime into a quiescent regime. As predicted by a dynamical model of hair-bundle mechanics, this boundary constitutes a Hopf bifurcation.

Bundled payments in orthopedic surgery.

PubMed

Bushnell, Brandon D

2015-02-01

As a result of reading this article, physicians should be able to: 1. Describe the concept of bundled payments and the potential applications of bundled payments in orthopedic surgery. 2. For specific situations, outline a clinical episode of care, determine the participants in a bundling situation, and define care protocols and pathways. 3. Recognize the importance of resource utilization management, quality outcome measurement, and combined economic-clinical value in determining the value of bundled payment arrangements. 4. Identify the implications of bundled payments for practicing orthopedists, as well as the legal issues and potential future directions of this increasingly popular alternative payment method. Bundled payments, the idea of paying a single price for a bundle of goods and services, is a financial concept familiar to most American consumers because examples appear in many industries. The idea of bundled payments has recently gained significant momentum as a financial model with the potential to decrease the significant current costs of health care. Orthopedic surgery as a field of medicine is uniquely positioned for success in an environment of bundled payments. This article reviews the history, logistics, and implications of the bundled payment model relative to orthopedic surgery. Copyright 2015, SLACK Incorporated.