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

Double-clad nuclear fuel safety rod

DOEpatents

McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

1984-01-01

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Double-clad nuclear-fuel safety rod

DOEpatents

McCarthy, W.H.; Atcheson, D.B.

1981-12-30

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Method and apparatus for reading lased bar codes on shiny-finished fuel rod cladding tubes

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

Goldenfield, M.P.; Lambert, D.V.

1990-10-02

This patent describes, in a nuclear fuel rod identification system, a method of reading a bar code etched directly on a surface of a nuclear fuel rod. It comprises: defining a pair of light diffuser surfaces adjacent one another but in oppositely inclined relation to a beam of light emitted from a light reader; positioning a fuel rod, having a cylindrical surface portion with a bar code etched directly thereon, relative to the light diffuser surfaces such that the surfaces are disposed adjacent to and in oppositely inclined relation along opposite sides of the fuel rod surface portion and themore » fuel rod surface portion is aligned with the beam of light emitted from the light reader; directing the beam of light on the bar code on fuel rod cylindrical surface portion such that the light is reflected therefrom onto one of the light diffuser surfaces; and receiving and reading the reflected light from the bar code via the one of the light diffuser surfaces to the light reader.« less

Benefits of barrier fuel on fuel cycle economics

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

Crowther, R.L.; Kunz, C.L.

1988-01-01

Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect ofmore » fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs.« less

Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

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

Siefken, L.J.

1999-01-01

Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from abovemore » on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown.« less

ACHIEVING THE REQUIRED COOLANT FLOW DISTRIBUTION FOR THE ACCELERATOR PRODUCTION OF TRITIUM (APT) TUNGSTEN NEUTRON SOURCE

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

D. SIEBE; K. PASAMEHMETOGLU

The Accelerator Production of Tritium neutron source consists of clad tungsten targets, which are concentric cylinders with a center rod. These targets are arranged in a matrix of tubes, producing a large number of parallel coolant paths. The coolant flow required to meet thermal-hydraulic design criteria varies with location. This paper describes the work performed to ensure an adequate coolant flow for each target for normal operation and residual heat-removal conditions.

Applicability of out-of-pile fretting wear tests to in-reactor fretting wear-induced failure time prediction

NASA Astrophysics Data System (ADS)

Kim, Kyu-Tae

2013-02-01

In order to investigate whether or not the grid-to-rod fretting wear-induced fuel failure will occur for newly developed spacer grid spring designs for the fuel lifetime, out-of-pile fretting wear tests with one or two fuel assemblies are to be performed. In this study, the out-of-pile fretting wear tests were performed in order to compare the potential for wear-induced fuel failure in two newly-developed, Korean PWR spacer grid designs. Lasting 20 days, the tests simulated maximum grid-to-rod gap conditions and the worst flow induced vibration effects that might take place over the fuel life time. The fuel rod perforation times calculated from the out-of-pile tests are greater than 1933 days for 2 μm oxidized fuel rods with a 100 μm grid-to-rod gap, whereas those estimated from in-reactor fretting wear failure database may be about in the range of between 60 and 100 days. This large discrepancy in fuel rod perforation may occur due to irradiation-induced cladding oxide microstructure changes on the one hand and a temperature gradient-induced hydrogen content profile across the cladding metal region on the other hand, which may accelerate brittleness in the grid-contacting cladding oxide and metal regions during the reactor operation. A three-phase grid-to-rod fretting wear model is proposed to simulate in-reactor fretting wear progress into the cladding, considering the microstructure changes of the cladding oxide and the hydrogen content profile across the cladding metal region combined with the temperature gradient. The out-of-pile tests cannot be directly applicable to the prediction of in-reactor fretting wear-induced cladding perforations but they can be used only for evaluating a relative wear resistance of one grid design against the other grid design.

Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

NASA Astrophysics Data System (ADS)

Wang, Hong; Wang, Jy-An John

2016-10-01

Behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending was studied. Tests were performed under load or moment control at 5 Hz. The surrogate rods fractured under moment amplitudes greater than 10.16 Nm with fatigue lives between 2.4 × 103 and 2.2 × 106 cycles. Fatigue response of Zry-4 cladding was characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition affect surrogate rod failure. Both debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. The effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective gauge length is effective in sensor spacing correction. The database developed and the understanding gained in this study can serve as input to analysis of SNF (spent nuclear fuel) vibration integrity.

Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

DOE PAGES

Wang, Hong; Wang, Jy-An John

2016-07-20

We studied behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending. Tests were performed under load or moment control at 5 Hz, and an empirical correlation was established between rod fatigue life and amplitude of the applied moment. Fatigue response of Zry-4 cladding was further characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment applied and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition all affect surrogate rod failure. Bonding/debonding of PPI/PCI and pellet fracturing contribute to surrogatemore » rod bending fatigue. Also, the effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective specimen gauge length is effective in sensor spacing correction. Finally, we developed the database and gained understanding in this study such that it will serve as input to analysis of SNF vibration integrity.« less

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

Jackson, Timothy D; Hollenbach, Daniel F; Shedlock, Daniel

Radiography by Selective Detection (RSD), was investigated for its ability to determine the presence and types of defects in a UO{sub 2} fuel rod surrounded by zirconium cladding. Images created using a Monte Carlo model compared favorably with actual X-ray backscatter images from mock fuel rods. A fuel rod was modeled as a rectangular parallelepiped with zirconium cladding, and pencil beam X-ray sources of 160 kVp (79 keV avg) and 480 kVp (218 keV avg) were generated using the Monte Carlo N-Particle Transport Code to attempt to image void and palladium (Pd) defects in the interior and on the surfacemore » of the fuel pellet. It was found that the 160 kVp spectrum was unable to detect the presence of interior defects, whereas the 480 kVp spectrum detected them with both the standard and the RSD backscatter methods, though the RSD method was very inefficient. It was also found that both energy spectra were able to detect void and Pd defects on the surface using both imaging methods. Additionally, two mock fuel rods were imaged using a backscatter X-ray imaging system, one consisting of hafnium pellets in a Zircaloy-4 cladding and the other consisting of steel pellets in a Zircalloy-4 cladding which was then encased in a steel cladding (a double encapsulation configuration employed in irradiation and experiments). It was found that the system was capable of detecting individual HfO{sub 2} pellets in a Zircaloy-4 cladding and may be capable of detecting individual steel pellets in the double-encapsulated sample. It is expected that the system would also be capable of detecting individual UO{sub 2} pellets in a Zircaloy-4 cladding, though no UO{sub 2} fuel rod was available for imaging.« less

Equations of state for crystalline zirconium iodide: The role of dispersion

NASA Astrophysics Data System (ADS)

Rossi, Matthew L.; Taylor, Christopher D.

2013-02-01

We present the first-principle equations of state of several zirconium iodides, ZrI2, ZrI3, and ZrI4, computed using density functional theory methods that apply various methods for introducing the dispersion correction. Iodides formed due to reaction of molecular or atomic iodine with zirconium and zircaloys are of particular interest due to their application to the cladding material used in the fabrication of nuclear fuel rods. Stress corrosion cracking (SCC), associated with fission product chemistry with the clad material, is a major concern in the life cycle of nuclear fuels, as many of the observed rod failures have occurred due to pellet-cladding chemical interactions (PCCI) [A. Atrens, G. Dannhäuser, G. Bäro, Stress-corrosion-cracking of zircaloy-4 cladding tubes, Journal of Nuclear Materials 126 (1984) 91-102; P. Rudling, R. Adamson, B. Cox, F. Garzarolli, A. Strasser, High burn-up fuel issues, Nuclear Engineering and Technology 40 (2008) 1-8]. A proper understanding of the physical properties of the corrosion products is, therefore, required for the development of a comprehensive SCC model. In this particular work, we emphasize that, while existing modeling techniques include methods to compute crystal structures and associated properties, it is important to capture intermolecular forces not traditionally included, such as van der Waals (dispersion) correction. Furthermore, crystal structures with stoichiometries favoring a high I:Zr ratio are found to be particularly sensitive, such that traditional density functional theory approaches that do not incorporate dispersion incorrectly predict significantly larger volumes of the lattice. This latter point is related to the diffuse nature of the iodide electron cloud.

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

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

Wirth, Brian; Terrani, Kurt A.; Sweet, Ryan T.

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling.more » As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.« less

Experimental verification of a theoretical model of an active cladding optical fiber fluorosensor

NASA Technical Reports Server (NTRS)

Albin, Sacharia; Briant, Alvin L.; Egalon, Claudio O.; Rogowski, Robert S.; Nankung, Juock S.

1993-01-01

Experiments were conducted to verify a theoretical model on the injection efficiency of sources in the cladding of an optical fiber. The theoretical results predicted an increase in the injection efficiency for higher differences in refractive indices between the core and cladding. The experimental apparatus used consisted of a glass rod 50 cm long, coated at one end with a thin film of fluorescent substance. The fluorescent substance was excited with side illumination, perpendicular to the rod axis, using a 476 nm Argon-ion laser. Part of the excited fluorescence was injected into the core and guided to a detector. The signal was measured for several different cladding refractive indices. The cladding consisted of sugar dissolved in water and the refractive index was changed by varying the sugar concentration in the solution. The results indicate that the power injected into the rod, due to evanescent wave injection, increases with the difference in refractive index which is in qualitative agreement with theory.

The COSIMA experiments and their verification, a data base for the validation of two phase flow computer codes

NASA Astrophysics Data System (ADS)

Class, G.; Meyder, R.; Stratmanns, E.

1985-12-01

The large data base for validation and development of computer codes for two-phase flow, generated at the COSIMA facility, is reviewed. The aim of COSIMA is to simulate the hydraulic, thermal, and mechanical conditions in the subchannel and the cladding of fuel rods in pressurized water reactors during the blowout phase of a loss of coolant accident. In terms of fuel rod behavior, it is found that during blowout under realistic conditions only small strains are reached. For cladding rupture extremely high rod internal pressures are necessary. The behavior of fuel rod simulators and the effect of thermocouples attached to the cladding outer surface are clarified. Calculations performed with the codes RELAP and DRUFAN show satisfactory agreement with experiments. This can be improved by updating the phase separation models in the codes.

Basic elements of light water reactor fuel rod design. [FUELROD code

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

Weisman, J.; Eckart, R.

1981-06-01

Basic design techniques and equations are presented to allow students to understand and perform preliminary fuel design for normal reactor conditions. Each of the important design considerations is presented and discussed in detail. These include the interaction between fuel pellets and cladding and the changes in fuel and cladding that occur during the operating lifetime of the fuel. A simple, student-oriented, fuel rod design computer program, called FUELROD, is described. The FUELROD program models the in-pile pellet cladding interaction and allows a realistic exploration of the effect of various design parameters. By use of FUELROD, the student can gain anmore » appreciation of the fuel rod design process. 34 refs.« less

Preliminary calculations related to the accident at Three Mile Island

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

Kirchner, W.L.; Stevenson, M.G.

This report discusses preliminary studies of the Three Mile Island Unit 2 (TMI-2) accident based on available methods and data. The work reported includes: (1) a TRAC base case calculation out to 3 hours into the accident sequence; (2) TRAC parametric calculations, these are the same as the base case except for a single hypothetical change in the system conditions, such as assuming the high pressure injection (HPI) system operated as designed rather than as in the accident; (3) fuel rod cladding failure, cladding oxidation due to zirconium metal-steam reactions, hydrogen release due to cladding oxidation, cladding ballooning, cladding embrittlement,more » and subsequent cladding breakup estimates based on TRAC calculated cladding temperatures and system pressures. Some conclusions of this work are: the TRAC base case accident calculation agrees very well with known system conditions to nearly 3 hours into the accident; the parametric calculations indicate that, loss-of-core cooling was most influenced by the throttling of High-Pressure Injection (HPI) flows, given the accident initiating events and the pressurizer electromagnetic-operated valve (EMOV) failing to close as designed; failure of nearly all the rods and gaseous fission product gas release from the failed rods is predicted to have occurred at about 2 hours and 30 minutes; cladding oxidation (zirconium-steam reaction) up to 3 hours resulted in the production of approximately 40 kilograms of hydrogen.« less

Compatibility studies on Mo-coating systems for nuclear fuel cladding applications

NASA Astrophysics Data System (ADS)

Koh, Huan Chin; Hosemann, Peter; Glaeser, Andreas M.; Cionea, Cristian

2017-12-01

To improve the safety factor of nuclear power plants in accident scenarios, molybdenum (Mo), with its high-temperature strength, is proposed as a potential fuel-cladding candidate. However, Mo undergoes rapid oxidation and sublimation at elevated temperatures in oxygen-rich environments. Thus, it is necessary to coat Mo with a protective layer. The diffusional interactions in two systems, namely, Zircaloy-2 (Zr2) on a Mo tube, and iron-chromium-aluminum (FeCrAl) on a Mo rod, were studied by aging coated Mo substrates in high vacuum at temperatures ranging from 650 °C to 1000° for 1000 h. The specimens were characterized using scanning electron microscopy (SEM), energy-dispersive spectrometry (EDS) and nanoindentation. In both systems, pores in the coating increased in size and number with increasing temperature over time, and cracks were also observed; intermetallic phases formed between the Mo and its coatings.

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

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

NEUTRONIC REACTOR CONTROL ROD AND METHOD OF FABRICATION

DOEpatents

Porembka, S.W. Jr.

1961-06-27

A reactor control rod formed from a compacted powder dispersion is patented. The rod consists of titanium sheathed with a cladding alloy. The cladding alloy contains 1.3% to 1.6% by weight of tin, 0.07% to 0.12% by weight of chromium, 0.04% to 0.08% by weight of nickel, 0.09% to 0.16% by weight of iron, carbon not exceeding 0.05%, less than 0.5% by weight of incidental impurities, and the balance zirconium.

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

NASA Astrophysics Data System (ADS)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

Strengthening effect of reduced graphene oxide in steel clad copper rod

NASA Astrophysics Data System (ADS)

Gao, Haitao; Liu, Xianghua; Ai, Zhengrong; Zhang, Shilong; Liu, Lizhong

2016-11-01

Reduced graphene oxide has been extensively used as reinforcing agent owing to their high mechanical properties. In this work, an attempt is made to synthesize steel clad copper rod reinforced with reduced graphene oxide (RGO) by the combination of powder-in-tube and intermediate annealing (IA). Experiments show that the Fe/RGO/Cu composites manifest better mechanical properties than Fe/Cu composites. In the process of groove rolling, RGO acts as effective binder, which can greatly improve the adhesive strength of copper scrap and two metals. Moreover, the strengthening effect of RGO is tightly related to its dispersion state. The RGO diffuses much more uniformly on the metallic substrate under the IA temperature of 1100 °C than 800 °C, which can be characterized by less deformation twins appearing at the interface of core copper and the formation of Fe-RGO-Cu transition belt at the bonding interface. In this case, the peak hardness, tensile strength and shear strength of Fe/RGO/Cu composites are 52 HV, 125 and 41 MPa higher than those of the Fe/Cu composites, respectively. The difference of strengthening effect and mechanisms of RGO under 800 and 1100 °C of IA are systematically discussed by referring to experimental results.

An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

NASA Astrophysics Data System (ADS)

Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

2018-01-01

Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

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

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

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

Sweet, Ryan; George, Nathan M.; Terrani, Kurt A.

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling themore » integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.« less

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

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

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

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

Hanan, N. A.; Matos, J. E.

At The request of the Czech Technical University in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. For core configurations C1 and C2, criticality calculations were done for cases with all control rodsmore » at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were done for the C1 core configuration. Finally the reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations.« less

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

Hannan, N. A.; Matos, J. E.; Stillman, J. A.

At the request of the Czech Technical University (CTU) in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. Fore core configurations C1 and C2, criticality calculations were done for cases with all controlmore » rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were doe for the C1 core configuration. The reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations. Finally, the reactivity feedback coefficients, the prompt neutron lifetime, and the total effective delay neutron fraction were calculated for each of the three cores.« less

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

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

Sridharan, Kumar; Mariani, Robert; Bai, Xianming

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins inmore » accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4) and silica (SiO 2) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.« less

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

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

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

DOE PAGES

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.; ...

2016-03-16

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

Modelling of LOCA Tests with the BISON Fuel Performance Code

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

Williamson, Richard L; Pastore, Giovanni; Novascone, Stephen Rhead

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculationsmore » are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.« less

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

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

Koyanagi, Takaaki; Petrie, Christian M.

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Officemore » of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high heat flux configurations have been assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. These rabbits contain a wide variety of specimens including monolith tubes, SiC fiber SiC matrix (SiC/SiC) composites, duplex specimens (inner composite, outer monolith), and specimens with a variety of metallic or ceramic coatings on the outer surface. The rabbits are targeted for insertion during HFIR cycle 475, which is scheduled for September 2017.« less

Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

NASA Astrophysics Data System (ADS)

Narukawa, Takafumi; Yamaguchi, Akira; Jang, Sunghyon; Amaya, Masaki

2018-02-01

For estimating fracture probability of fuel cladding tube under loss-of-coolant accident conditions of light-water-reactors, laboratory-scale integral thermal shock tests were conducted on non-irradiated Zircaloy-4 cladding tube specimens. Then, the obtained binary data with respect to fracture or non-fracture of the cladding tube specimen were analyzed statistically. A method to obtain the fracture probability curve as a function of equivalent cladding reacted (ECR) was proposed using Bayesian inference for generalized linear models: probit, logit, and log-probit models. Then, model selection was performed in terms of physical characteristics and information criteria, a widely applicable information criterion and a widely applicable Bayesian information criterion. As a result, it was clarified that the log-probit model was the best among the three models to estimate the fracture probability in terms of the degree of prediction accuracy for both next data to be obtained and the true model. Using the log-probit model, it was shown that 20% ECR corresponded to a 5% probability level with a 95% confidence of fracture of the cladding tube specimens.

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

Wang, Hong; Wang, Jy-An John

We studied behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending. Tests were performed under load or moment control at 5 Hz, and an empirical correlation was established between rod fatigue life and amplitude of the applied moment. Fatigue response of Zry-4 cladding was further characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment applied and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition all affect surrogate rod failure. Bonding/debonding of PPI/PCI and pellet fracturing contribute to surrogatemore » rod bending fatigue. Also, the effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective specimen gauge length is effective in sensor spacing correction. Finally, we developed the database and gained understanding in this study such that it will serve as input to analysis of SNF vibration integrity.« less

3D modeling of missing pellet surface defects in BWR fuel

DOE PAGES

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.; ...

2016-07-26

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

Fabrication of stainless steel clad tubing. [gas pressure bonding

NASA Technical Reports Server (NTRS)

Kovach, C. W.

1978-01-01

The feasibility of producing stainless steel clad carbon steel tubing by a gas pressure bonding process was evaluated. Such a tube product could provide substantial chromium savings over monolithic stainless tubing in the event of a serious chromium shortage. The process consists of the initial assembly of three component tubesets from conventionally produced tubing, the formation of a strong metallurgical bond between the three components by gas pressure bonding, and conventional cold draw and anneal processing to final size. The quality of the tubes produced was excellent from the standpoint of bond strength, mechanical, and forming properties. The only significant quality problem encountered was carburization of the stainless clad by the carbon steel core which can be overcome by further refinement through at least three different approaches. The estimated cost of clad tubing produced by this process is greater than that for monolithic stainless tubing, but not so high as to make the process impractical as a chromium conservation method.

Thermal hydraulic design and decay heat removal of a solid target for a spallation neutron source

NASA Astrophysics Data System (ADS)

Takenaka, N.; Nio, D.; Kiyanagi, Y.; Mishima, K.; Kawai, M.; Furusaka, M.

2005-08-01

Thermal hydraulic design and thermal stress calculations were conducted for a water-cooled solid target irradiated by a MW-class proton beam for a spallation neutron source. Plate type and rod bundle type targets were examined. The thickness of the plate and the diameter of the rod were determined based on the maximum and the wall surface temperature. The thermal stress distributions were calculated by a finite element method (FEM). The neutronics performance of the target is roughly proportional to its average density. The averaged densities of the designed targets were calculated for tungsten plates, tantalum clad tungsten plates, tungsten rods sheathed by tantalum and Zircaloy and they were compared with mercury density. It was shown that the averaged density was highest for the tungsten plates and was high for the tantalum cladding tungsten plates, the tungsten rods sheathed by tantalum and Zircaloy in order. They were higher than or equal to that of mercury for the 1 2 MW proton beams. Tungsten target without the cladding or the sheath is not practical due to corrosion by water under irradiation condition. Therefore, the tantalum cladding tungsten plate already made successfully by HIP and the sheathed tungsten rod are the candidate of high performance solid targets. The decay heat of each target was calculated. It was low enough low compared to that of ISIS for the target without tantalum but was about four times as high as that of ISIS when the thickness of the tantalum cladding was 0.5 mm. Heat removal methods of the decay heat with tantalum were examined. It was shown that a special cooling system was required for the target exchange when tantalum was used for the target. It was concluded that the tungsten rod target sheathed with stainless steel or Zircaloy was the most reliable from the safety considerations and had similar neutronics performance to that of mercury.

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.

The effect of silicon on the interaction between metallic uranium and aluminum: A 50 year long diffusion experiment

NASA Astrophysics Data System (ADS)

Leenaers, A.; Detavernier, C.; Van den Berghe, S.

2008-11-01

The core of the BR1 research reactor at SCK•CEN, Mol (Belgium) has a graphite matrix loaded with fuel rods consisting of a natural uranium slug in aluminum cladding. The BR1 reactor has been in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, some of the fuel rods have been examined. Fabrication reports indicate that a so-called AlSi bonding layer and an U(Al,Si) 3 anti-diffusion layer on the natural uranium fuel slug were applied to limit the interaction between the uranium fuel and aluminum cladding. The microstructure of the fuel, bonding and anti-diffusion layer and cladding were analysed using optical microscopy, scanning electron microscopy and electron microprobe analysis. It was found that the AlSi bonding layer does provide a tight bond between fuel and cladding but that it is a thin USi layer that acts as effective anti-diffusion layer and not the intended U(Al,Si) 3 layer.

Design of cladding rods-assisted depressed-core few-mode fibers with improved modal spacing

NASA Astrophysics Data System (ADS)

Han, Jiawei; Zhang, Jie

2018-03-01

This paper investigates the design details of cladding rods-assisted (CRA) depressed-core (DC) few-mode fibers (FMFs) that feature more equally spaced linearly polarized (LP) modal effective indices, suitable for high-spatial-density weakly-coupled mode-division multiplexing systems. The influences of the index profile of cladding rods on LP mode-resolved effective index, bending sensitivity, and effective area Aeff, are numerically described. Based on the design considerations of LP modal Aeff-dependent spatial efficiency and LP modal bending loss-dependent robustness, the small LP21-LP02 and LP22-LP03 modal spacing limitations, encountered in state-of-the-art weakly-coupled step-index FMFs, have been substantially improved by at least 25%. In addition, the proposed CRA DC FMFs also show sufficiently large effective areas (in excess of 110 μm2) for all guided LP modes, which are expected to exhibit good nonlinear performance.

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

Densities ............................................................................................................ 21 2.3 Fuel Mass (Core Total...70 7.1 Geometry, Material Density, and Mass Summary for All Cores...21 Table 3: Fuel Rod Masses for Different Clads

iMAST FY2007 Annual Report

DTIC Science & Technology

2007-01-01

laser cladding techniques. Customer: COMSUBPAC NAVSEASYSCOM Pearl Harbor Naval Shipyard...motion device to laser clad and re-dimension the affected tube to original specifications. Benefits: ° Reduce life cycle costs of tube...coatings via cold gas dynamic spraying and EB–PVD ° Spray-formed HT aluminum alloys ° Localized laser HT and cladding for wear

Synthesis, characterization and processing of active rare earth-doped chalcohalide glasses

NASA Astrophysics Data System (ADS)

Debari, Roberto Mauro

Applications for infrared-transmitting non-oxide glass fibers span a broad range of topics. They can be used in the military, the medical field, telecommunications, and even in agriculture. Rare earth ions are used as dopants in these glasses in order to stimulate emissions in the infrared spectral region. In order to extend the host glass transmission further into the infrared, selenium atoms were substituted for sulfur in the established Ge-S-I chalcohalide glass system and the fundamental properties of these latter glasses were explored. Over 30 different compositions in the Ge-Se-I glass system were investigated as to their thermal and optical properties. The resulting optimum host with a composition of Ge15Se80I5 has a broad transmission range from 0.7 mum to 17.0 mum and a high working range over 145°C. The host glass also exhibited a Tg of 125°C, making rotational casting of a cladding tube for rod-and-tube fiberization a possibility. The base glass was doped with 1000 to 4000 ppm/wt of erbium, dysprosium, or neodymium. When doped with Er3+-ions, absorptions at 1.54 mum and 3.42 mum were observed. Nd3+-doping resulted in an absorption peak near 4.24 mum and Dy3+ ions caused absorption at 1.30 mum. Fluorescence emissions were found for neodymium at 1.396 mum with a FWHM of 74 nm, and for dysprosium at 1.145 mum with a FWHM of 75 nm, at 1.360 mum with a FWHM of 98 rim and at 1.674 mum with a FWHM of 60 nm. High optical quality tubes of the host glass could be formed using rotational casting in silica ampoules. Glass tubes, 4 to 6 cm long with a 1 cm outer diameter and a tailored inner-hole diameter ranging from 0.4 to 0.6 cm could be synthesized by this process with excellent dimensional tolerances around the circumference as well as along the length. A preform of this size provided 25 continuous meters of unclad fiber with diameters ranging from 140 to 200 mum. A UV-curable acrylate cladding was applied via an external coating cup. An x-ray analysis of the resulting fiber verified the constituents of the fiber. Due to tradeoffs between thermal properties, optical properties and rare earth solubility, the Ge-Se-I glass system must still be optimized prior to use as an active fiber device. Nevertheless, the viability of this host system has been demonstrated in this investigation. Some very promising advantages to adding halides to chalcogenide glass systems have been confirmed, including the tailoring of glass transition temperatures, enhancement of rare earth solubility, expanded fluorescence emissions in the IR, and suppression of some impurity absorption bands. Also, the potential for rod-and-tube fiberization utilizing the rotational casting method for tube synthesis has been established along with its resulting pristine core-clad interface. This research provides a foundation for active fiber device applications in the 2 to 10 mum spectral region.

Nuclear reactor fuel element having improved heat transfer

DOEpatents

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

1982-03-03

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

Cladding material, tube including such cladding material and methods of forming the same

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

Garnier, John E.; Griffith, George W.

A multi-layered cladding material including a ceramic matrix composite and a metallic material, and a tube formed from the cladding material. The metallic material forms an inner liner of the tube and enables hermetic sealing of thereof. The metallic material at ends of the tube may be exposed and have an increased thickness enabling end cap welding. The metallic material may, optionally, be formed to infiltrate voids in the ceramic matrix composite, the ceramic matrix composite encapsulated by the metallic material. The ceramic matrix composite includes a fiber reinforcement and provides increased mechanical strength, stiffness, thermal shock resistance and highmore » temperature load capacity to the metallic material of the inner liner. The tube may be used as a containment vessel for nuclear fuel used in a nuclear power plant or other reactor. Methods for forming the tube comprising the ceramic matrix composite and the metallic material are also disclosed.« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1980-04-29

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

Methods and codes for neutronic calculations of the MARIA research reactor.

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

Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

2002-02-18

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

Reversible Bending Fatigue Test System for Investigating Vibration Integrity of Spent Nuclear Fuel during Transportation

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

Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading during road or rail shipment. Oak Ridge National Laboratory (ORNL) has been developing testing capabilities that can be used to improve the understanding of the impacts on SNF integrity due to vibration loading, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in themore » area of safety and security of spent nuclear fuel storage and transport operations. The ORNL developed test system can perform reversible-bending fatigue testing to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The testing apparatus is also designed to meet the challenges of hot-cell operation, including remote installation and detachment of the SNF test specimen, in-situ test specimen deformation measurement, and implementation of a driving system suitable for use in a hot cell. The system contains a U-frame set-up equipped with uniquely designed grip rigs, to protect SNF rod and to ensure valid test results, and use of 3 specially designed LVDTs to obtain the in-situ curvature measurement. A variety of surrogate test rods have been used to develop and calibrate the test system as well as in performing a series of systematic cyclic fatigue tests. The surrogate rods include stainless steel (SS) cladding, SS cladding with cast epoxy, and SS cladding with alumina pellets inserts simulating fuel pellets. Testing to date has shown that the interface bonding between the SS cladding and the alumina pellets has a significant impact on the bending response of the test rods as well as their fatigue strength. The failure behaviors observed from tested surrogate rods provides a fundamental understanding of the underlying failure mechanisms of the SNF surrogate rod under vibration which has not been achieved previously. The newly developed device is scheduled to be installed in the hot-cell in summer 2013 to test high burnup SNF.« less

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

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

None

1960-10-31

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

Spent fuel behavior under abnormal thermal transients during dry storage

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

Stahl, D.; Landow, M.P.; Burian, R.J.

1986-01-01

This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment wasmore » heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.« less

Fission gas release restrictor for breached fuel rod

DOEpatents

Kadambi, N. Prasad; Tilbrook, Roger W.; Spencer, Daniel R.; Schwallie, Ambrose L.

1986-01-01

In the event of a breach in the cladding of a rod in an operating liquid metal fast breeder reactor, the rapid release of high-pressure gas from the fission gas plenum may result in a gas blanketing of the breached rod and rods adjacent thereto which impairs the heat transfer to the liquid metal coolant. In order to control the release rate of fission gas in the event of a breached rod, the substantial portion of the conventional fission gas plenum is formed as a gas bottle means which includes a gas pervious means in a small portion thereof. During normal reactor operation, as the fission gas pressure gradually increases, the gas pressure interiorly of and exteriorly of the gas bottle means equalizes. In the event of a breach in the cladding, the gas pervious means in the gas bottle means constitutes a sufficient restriction to the rapid flow of gas therethrough that under maximum design pressure differential conditions, the fission gas flow through the breach will not significantly reduce the heat transfer from the affected rod and adjacent rods to the liquid metal heat transfer fluid flowing therebetween.

Corrosion inhibition of steam generator tubesheet by Alloy 690 cladding in secondary side environments

NASA Astrophysics Data System (ADS)

Hur, Do Haeng; Choi, Myung Sik; Lee, Deok Hyun; Han, Jung Ho; Shim, Hee Sang

2013-11-01

Denting is a phenomenon that a steam generator tube is distorted by a volume expansion of corrosion products of the tube support and tubesheet materials adjacent to the tube. Although denting has been mitigated by a modification of the design and material of the tube support structures, it has been an inevitable concern in the crevice region of the top of tubesheet. This paper provides a new technology to prevent denting by cladding the secondary surface of the tubesheet with a corrosion resistant material. In this study, Alloy 690 material was cladded onto the surface of an SA508 tubesheet to a thickness of about 9 mm. The corrosion rates of the original SA508 tubesheet and the Alloy 690 clad material were measured in acidic and alkaline simulated environments. Using Alloy 690 cladding, the corrosion rate of the tubesheet within a magnetite sludge pile decreased by a factor of 680 in 0.1 M NiCl2 solution at 300 °C, and by a factor of 58 in 2 M NaOH solution at 315 °C. This means that denting can drastically be prevented by cladding the secondary tubesheet surface with corrosion resistant materials.

Method and means of packaging nuclear fuel rods for handling

DOEpatents

Adam, Milton F.

1979-01-01

Nuclear fuel rods, especially spent nuclear fuel rods that may show physical distortion, are encased within a metallic enclosing structure by forming a tube about the fuel rod. The tube has previously been rolled to form an overlapping tubular structure and then unrolled and coiled about an axis perpendicular to the tube. The fuel rod is inserted into the tube as the rolled tube is removed from a coiled strip and allowed to reassume its tubular shape about the fuel rod. Rollers support the coiled strip in an open position as the coiled strip is uncoiled and allowed to roll about the fuel rod.

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

LIGHT WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Christy, R.F.; Weinberg, A.M.

1957-09-17

A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

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

McHugh, Kevin M; Garnier, John E; Sergey Rashkeev

2013-01-01

Stability of cladding materials under off-normal conditions is an important issue for the safe operation of light water nuclear reactors. Metals, ceramics, and metal/ceramic composites are being investigated as substitutes for traditional zirconium-based cladding. To support down-selection of these advanced materials and designs, a test apparatus was constructed to study the onset and evolution of cladding oxidation, and deformation behavior of cladding materials, under loss-of-coolant accident scenarios. Preliminary oxidation tests were conducted in dry oxygen and in saturated steam/air environments at 1000OC. Tube samples of Zr-702, Zr-702 reinforced with 1 ply of a ß-SiC CMC overbraid, and sintered a-SiC weremore » tested. Samples were induction heated by coupling to a molybdenum susceptor inside the tubes. The deformation behavior of He-pressurized tubes of Zr-702 and SiC CMC-reinforced Zr-702, heated to rupture, was also examined.« less

Fluorescent discharge lamp

NASA Technical Reports Server (NTRS)

Mukai, E.; Otsuka, H.; Nomi, K.; Honmo, I.

1982-01-01

A rapidly illuminating fluorescent lamp 1,200 mm long and 32.5 mm in diameter with an interior conducting strip which is compatible with conventional fixtures and ballasts is described. The fluorescent lamp is composed of a linear glass tube, electrodes sealed at both ends, mercury and raregas sealed in the glass tube, a fluorescent substance clad on the inner walls of the glass tube, and a clad conducting strip extending the entire length of the glass tube in the axial direction on the inner surface of the tube.

Development of U-frame bending system for studying the vibration integrity of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Wang, Hong; Wang, Jy-An John; Tan, Ting; Jiang, Hao; Cox, Thomas S.; Howard, Rob L.; Bevard, Bruce B.; Flanagan, Michelle

2013-09-01

A bending fatigue system developed to evaluate the response of spent nuclear fuel rods to vibration loads is presented. A U-frame testing setup is used for imposing bending loads on the fuel rod specimen. The U-frame setup consists of two rigid arms, side connecting plates to the rigid arms, and linkages to a universal testing machine. The test specimen's curvature is obtained through a three-point deflection measurement method. The tests using surrogate specimens with stainless steel cladding revealed increased flexural rigidity under unidirectional cyclic bending, significant effect of cladding-pellets bonding on the response of surrogate rods, and substantial cyclic softening in reverse bending mode. These phenomena may cast light on the expected response of a spent nuclear fuel rod. The developed U-frame system is thus verified and demonstrated to be ready for further pursuit in hot-cell tests.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

A Multi-Stage Wear Model for Grid-to-Rod Fretting of Nuclear Fuel Rods

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

Blau, Peter Julian

The wear of fuel rod cladding against the supporting structures in the cores of pressurized water nuclear reactors (PWRs) is an important and potentially costly tribological issue. Grid-to-rod fretting (GTRF), as it is known, involves not only time-varying contact conditions, but also elevated temperatures, flowing hot water, aqueous tribo-corrosion, and the embrittling effects of neutron fluences. The multi-stage, closed-form analytical model described in this paper relies on published out-of-reactor wear and corrosion data and a set of simplifying assumptions to portray the conversion of frictional work into wear depth. The cladding material of interest is a zirconium-based alloy called Zircaloy-4,more » and the grid support is made of a harder and more wear-resistant material. Focus is on the wear of the cladding. The model involves an incubation stage, a surface oxide wear stage, and a base alloy wear stage. The wear coefficient, which is a measure of the efficiency of conversion of frictional work into wear damage, can change to reflect the evolving metallurgical condition of the alloy. Wear coefficients for Zircaloy-4 and for a polyphase zirconia layer were back-calculated for a range of times required to wear to a critical depth. Inputs for the model, like the friction coefficient, are taken from the tribology literature in lieu of in-reactor tribological data. Concepts of classical fretting were used as a basis, but are modified to enable the model to accommodate the complexities of the PWR environment. Factors like grid spring relaxation, pre-oxidation of the cladding, multiple oxide phases, gap formation, impact, and hydrogen embrittlement are part of the problem definition but uncertainties in their relative roles limits the ability to validate the model. Sample calculations of wear depth versus time in the cladding illustrate how GTRF wear might occur in a discontinuous fashion during months-long reactor operating cycles. A means to account for grid/rod gaps and repetitive impact effects on GTRF wear is proposed« less

Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

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

Marshall, Margaret A.; Bess, John D.

2015-02-01

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965 have been evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study “power plants for the production of electrical power in space vehicles.” The third configuration in this series of experiments was a beryllium-reflected assembly of stainless-steel-clad, highly enriched uranium (HEU)-O 2 fuel mockup of a potassium-cooledmore » space power reactor. Reactivity measurements cadmium ratio spectral measurements and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The fuel tube effect and neutron moderating and absorbing material worth measurements are the focus of this paper. Additionally, a measurement of the worth of potassium filling the core region was performed but has not yet been evaluated Pellets of 93.15 wt.% enriched uranium dioxide (UO 2) were stacked in 30.48 cm tall stainless steel fuel tubes (0.3 cm tall end caps). Each fuel tube had 26 pellets with a total mass of 295.8 g UO 2 per tube. 253 tubes were arranged in 1.506-cm triangular lattice. An additional 7-tube cluster critical configuration was also measured but not used for any physics measurements. The core was surrounded on all side by a beryllium reflector. The fuel effect worths were measured by removing fuel tubes at various radius. An accident scenario was also simulated by moving outward twenty fuel rods from the periphery of the core so they were touching the core tank. The change in the system reactivity when the fuel tube(s) were removed/moved compared with the base configuration was the worth of the fuel tubes or accident scenario. The worth of neutron absorbing and moderating materials was measured by inserting material rods into the core at regular intervals or placing lids at the top of the core tank. Stainless steel 347, tungsten, niobium, polyethylene, graphite, boron carbide, aluminum and cadmium rods and/or lid worths were all measured. The change in the system reactivity when a material was inserted into the core is the worth of the material.« less

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

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

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

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

DOE PAGES

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

2016-09-26

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

A comparative study of the mechanical properties and the behavior of carbon and boron in stainless steel cladding tubes fabricated by PM HIP and traditional technologies

NASA Astrophysics Data System (ADS)

Shulga, A. V.

2013-03-01

The ring tensile test method was optimized and successfully used to obtain precise data for specimens of the cladding tubes of AISI type 316 austenitic stainless steels and ferritic-martensitic stainless steel. The positive modifications in the tensile properties of the stainless steel cladding tubes fabricated by powder metallurgy and hot isostatic pressing of melt atomized powders (PM HIP) when compared with the cladding tubes produced by traditional technology were found. Presently, PM HIP is also used in the fabrication of oxide dispersion strengthened (ODS) ferritic-martensitic steels. The high degree of homogeneity of the distribution of carbon and boron as well the high dispersivity of the phase-structure elements in the specimens manufactured via PM HIP were determined by direct autoradiography methods. These results correlate well with the increase of the tensile properties of the specimens produced by PM HIP technology.

The effect of stress state on zirconium hydride reorientation

NASA Astrophysics Data System (ADS)

Cinbiz, Mahmut Nedim

Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multiaxial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading. Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation. To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to "near-equibiaxial" tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by correlating the finite element stress-state results with the spatial distribution of hydride microstructures observed within the optical micrographs for each sample. Experiments showed that the hydride reorientation was enhanced as the stress biaxiality increased. The threshold stress decreased from 150 MPa to 80 MPa when stress biaxiality ratio increased from uniaxial tension to near-equibiaxial tension. This behavior was also predicted by classical nucleation theory based on the Gibbs free energy of transformation being assisted by the far-field stress. An analysis of in situ X-ray diffraction data obtained during a thermo-mechanical cycle typical of vacuum drying showed a complex lattice-spacing behavior of the hydride phase during the dissolution and precipitation. The in-plane hydrides showed bilinear lattice expansion during heating with the intrinsic thermal expansion rate of the hydrides being observed only at elevated temperatures as they dissolve. For radial hydrides that precipitate during cooling under stress, the spacing of the close-packed {111} planes oriented normal to the maximum applied stress was permanently higher than the corresponding {111} plane spacing in the other directions. This behavior is believed to be a result of a complex stress state within the precipitating plate-like hydrides that induces a strain component within the hydrides normal to its "plate" face (i.e., the applied stress direction) that exceeds the lattice spacing strains in the other directions. During heat-up, the lattice spacing of these same "plate" planes actually contract due to the reversion of the stress state within the plate-like hydrides as they dissolve. The presence of radial hydrides and their connectivity with in-plane hydrides was shown to increase the ductile-to-brittle transition temperature during tensile testing. This behavior can be understood in terms of the role of radial hydrides in promoting the initiation of a long crack that subsequently propagates under fracture mechanics conditions. Finally, the d-spacing of irradiated Zircaloy-4 and M5 cladding tubes was measured at room temperature and compared to that of unirradiated samples.

Novel Accident-Tolerant Fuel Meat and Cladding

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

Robert D. Mariani; Pavel G Medvedev; Douglas L Porter

A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas releasemore » and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.« less

Defect scriber

DOEpatents

Russell, Harold C.

1979-01-01

This disclosure describes a device for repeatably scribing a V-shaped scratch having sharply defined dimensions on the interior surface of a nuclear reactor fuel rod tube. A cutting tool having a V-shaped cutting tip is supported within the fuel rod tube so that the V-shaped cutting tip can be pivoted about an axis and scribe a scratch on the interior surface of the fuel rod tube. Lengthwise the scratch runs parallel to a line drawn through the axis of the fuel rod tube and is in the shape of an arc, and widthwise the scratch is V-shaped. This shape is used because the dimensions of the scratch can be plugged into appropriate formulas to calculate stress intensity of cracks in fuel rod tubes. Since the fuel rod tubes which are to be scribed may be radioactive, the scratching assembly is designed for use in a fixture which allows it to be operated in a cave by remote control handling devices.

CTF Preprocessor User's Manual

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

Avramova, Maria; Salko, Robert K.

2016-05-26

This document describes how a user should go about using the CTF pre- processor tool to create an input deck for modeling rod-bundle geometry in CTF. The tool was designed to generate input decks in a quick and less error-prone manner for CTF. The pre-processor is a completely independent utility, written in Fortran, that takes a reduced amount of input from the user. The information that the user must supply is basic information on bundle geometry, such as rod pitch, clad thickness, and axial location of spacer grids--the pre-processor takes this basic information and determines channel placement and connection informationmore » to be written to the input deck, which is the most time-consuming and error-prone segment of creating a deck. Creation of the model is also more intuitive, as the user can specify assembly and water-tube placement using visual maps instead of having to place them by determining channel/channel and rod/channel connections. As an example of the benefit of the pre-processor, a quarter-core model that contains 500,000 scalar-mesh cells was read into CTF from an input deck containing 200,000 lines of data. This 200,000 line input deck was produced automatically from a set of pre-processor decks that contained only 300 lines of data.« less

Cladding For Transversely-Pumped Laser Rod

NASA Technical Reports Server (NTRS)

Byer, Robert L.; Fan, Tso Yee

1989-01-01

Combination of suitable dimensioning and cladding of neodymium:yttrium aluminum garnet of similar solid-state laser provides for more efficient utilization of transversely-incident pump light from diode lasers. New design overcomes some of limitations of longitudinal- and older transverse-pumping concepts and promotes operation at higher output powers in TEM00 mode.

Chemical vapor deposition of Mo tubes for fuel cladding applications

DOE PAGES

Beaux, Miles F.; Vodnik, Douglas R.; Peterson, Reuben J.; ...

2018-01-31

In this study, chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wallmore » thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. Finally, a path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system.« less

Chemical vapor deposition of Mo tubes for fuel cladding applications

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

Beaux, Miles F.; Vodnik, Douglas R.; Peterson, Reuben J.

In this study, chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wallmore » thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. Finally, a path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system.« less

Mechanical and degradation properties of biodegradable Mg strengthened poly-lactic acid composite through plastic injection molding.

PubMed

Butt, Muhammad Shoaib; Bai, Jing; Wan, Xiaofeng; Chu, Chenglin; Xue, Feng; Ding, Hongyan; Zhou, Guanghong

2017-01-01

Full biodegradable magnesium alloy (AZ31) strengthened poly-lactic acid (PLA) composite rods for potential application for bone fracture fixation were prepared by plastic injection process in this work. Their surface/interfacial morphologies, mechanical properties and vitro degradation were studied. In comparison with untreated Mg rod, porous MgO ceramic coating on Mg surface formed by Anodizing (AO) and micro-arc-oxidation (MAO)treatment can significantly improve the interfacial binding between outer PLA cladding and inner Mg rod due to the micro-anchoring action, leading to better mechanical properties and degradation performance of the composite rods.With prolonging immersion time in simulated body fluid (SBF) solution until 8weeks, the MgO porous coating were corroded gradually, along with the disappearance of original pores and the formation of a relatively smooth surface. This resulted in a rapidly reduction in mechanical properties for corresponding composite rods owing to the weakening of interfacial binding capacity. The present results indicated that this new PLA-clad Mg composite rods show good potential biomedical applications for implants and instruments of orthopedic inner fixation. Copyright © 2016 Elsevier B.V. All rights reserved.

Glass-ceramic optical fiber containing Ba2TiSi2O8 nanocrystals for frequency conversion of lasers

NASA Astrophysics Data System (ADS)

Fang, Zaijin; Xiao, Xusheng; Wang, Xin; Ma, Zhijun; Lewis, Elfed; Farrell, Gerald; Wang, Pengfei; Ren, Jing; Guo, Haitao; Qiu, Jianrong

2017-03-01

A glass-ceramic optical fiber containing Ba2TiSi2O8 nanocrystals fabricated using a novel combination of the melt-in-tube method and successive heat treatment is reported for the first time. For the melt-in-tube method, fibers act as a precursor at the drawing temperature for which the cladding glass is softened while the core glass is melted. It is demonstrated experimentally that following heat treatment, Ba2TiSi2O8 nanocrystals with diameters below 10 nm are evenly distributed throughout the fiber core. Comparing to the conventional rod-in-tube method, the melt-in-tube method is superior in terms of controllability of crystallization to allow for the fabrication of low loss glass-ceramic fibers. When irradiated using a 1030 nm femtosecond laser, an enhanced green emission at a wavelength of 515 nm is observed in the glass-ceramic fiber, which demonstrates second harmonic generation of a laser action in the fabricated glass-ceramic fibers. Therefore, this new glass-ceramic fiber not only provides a highly promising development for frequency conversion of lasers in all optical fiber based networks, but the melt-in-tube fabrication method also offers excellent opportunities for fabricating a wide range of novel glass-ceramic optical fibers for multiple future applications including fiber telecommunications and lasers.

Glass-ceramic optical fiber containing Ba2TiSi2O8 nanocrystals for frequency conversion of lasers

PubMed Central

Fang, Zaijin; Xiao, Xusheng; Wang, Xin; Ma, Zhijun; Lewis, Elfed; Farrell, Gerald; Wang, Pengfei; Ren, Jing; Guo, Haitao; Qiu, Jianrong

2017-01-01

A glass-ceramic optical fiber containing Ba2TiSi2O8 nanocrystals fabricated using a novel combination of the melt-in-tube method and successive heat treatment is reported for the first time. For the melt-in-tube method, fibers act as a precursor at the drawing temperature for which the cladding glass is softened while the core glass is melted. It is demonstrated experimentally that following heat treatment, Ba2TiSi2O8 nanocrystals with diameters below 10 nm are evenly distributed throughout the fiber core. Comparing to the conventional rod-in-tube method, the melt-in-tube method is superior in terms of controllability of crystallization to allow for the fabrication of low loss glass-ceramic fibers. When irradiated using a 1030 nm femtosecond laser, an enhanced green emission at a wavelength of 515 nm is observed in the glass-ceramic fiber, which demonstrates second harmonic generation of a laser action in the fabricated glass-ceramic fibers. Therefore, this new glass-ceramic fiber not only provides a highly promising development for frequency conversion of lasers in all optical fiber based networks, but the melt-in-tube fabrication method also offers excellent opportunities for fabricating a wide range of novel glass-ceramic optical fibers for multiple future applications including fiber telecommunications and lasers. PMID:28358045

Glass-ceramic optical fiber containing Ba2TiSi2O8 nanocrystals for frequency conversion of lasers.

PubMed

Fang, Zaijin; Xiao, Xusheng; Wang, Xin; Ma, Zhijun; Lewis, Elfed; Farrell, Gerald; Wang, Pengfei; Ren, Jing; Guo, Haitao; Qiu, Jianrong

2017-03-30

A glass-ceramic optical fiber containing Ba 2 TiSi 2 O 8 nanocrystals fabricated using a novel combination of the melt-in-tube method and successive heat treatment is reported for the first time. For the melt-in-tube method, fibers act as a precursor at the drawing temperature for which the cladding glass is softened while the core glass is melted. It is demonstrated experimentally that following heat treatment, Ba 2 TiSi 2 O 8 nanocrystals with diameters below 10 nm are evenly distributed throughout the fiber core. Comparing to the conventional rod-in-tube method, the melt-in-tube method is superior in terms of controllability of crystallization to allow for the fabrication of low loss glass-ceramic fibers. When irradiated using a 1030 nm femtosecond laser, an enhanced green emission at a wavelength of 515 nm is observed in the glass-ceramic fiber, which demonstrates second harmonic generation of a laser action in the fabricated glass-ceramic fibers. Therefore, this new glass-ceramic fiber not only provides a highly promising development for frequency conversion of lasers in all optical fiber based networks, but the melt-in-tube fabrication method also offers excellent opportunities for fabricating a wide range of novel glass-ceramic optical fibers for multiple future applications including fiber telecommunications and lasers.

Electroslag Strip Cladding of Steam Generators With Alloy 690

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

Consonni, M.; Maggioni, F.; Brioschi, F.

2006-07-01

The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations' steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layermore » leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallization effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding. (authors)« less

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.

Hybrid nuclear reactor grey rod to obtain required reactivity worth

DOEpatents

Miller, John V.; Carlson, William R.; Yarbrough, Michael B.

1991-01-01

Hybrid nuclear reactor grey rods are described, wherein geometric combinations of relatively weak neutron absorber materials such as stainless steel, zirconium or INCONEL, and relatively strong neutron absorber materials, such as hafnium, silver-indium cadmium and boron carbide, are used to obtain the reactivity worths required to reach zero boron change load follow. One embodiment includes a grey rod which has combinations of weak and strong neutron absorber pellets in a stainless steel cladding. The respective pellets can be of differing heights. A second embodiment includes a grey rod with a relatively thick stainless steel cladding receiving relatively strong neutron absorber pellets only. A third embodiment includes annular relatively weak netron absorber pellets with a smaller diameter pellet of relatively strong absorber material contained within the aperture of each relatively weak absorber pellet. The fourth embodiment includes pellets made of a homogeneous alloy of hafnium and a relatively weak absorber material, with the percentage of hafnium chosen to obtain the desired reactivity worth.

Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

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

Blau, P. J.; Qu, J.; Lu, R.

One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes themore » basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.« less

Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

DOE PAGES

Blau, P. J.; Qu, J.; Lu, R.

2016-09-21

One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes themore » basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.« less

The honey insertion cladding to improve the sensitivity of temperature polymer optical fiber sensor

NASA Astrophysics Data System (ADS)

Arwani, M.; Kuswanto, H.

2018-04-01

The sensitivity of temperature polymer optical fiber (POF) sensor has been studied. Part of cladding (9 cm) was substituted with honey. Polymer cladding was stripped mechanically and the honey inserted into the tube. Plastic gel closed the two end sides of the tubes. The optical power output was detected by Optical Power Meter (OPM). Honey cladding and temperature changing effect to the internal reflection and optical fiber output intensity. Highest output intensity changing at 20°C was shown by optical fiber coated by longan honey as cladding. The range of 10-50° C, as the rise of surroundings temperature, the attenuation was getting smaller. Best sensitivity was fiber with sensing part coated by Longan honey. Best linearity was sensing fiber with sensing part coated by Pracimantoro honey.

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

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

McCulloch, R.W.; MacPherson, R.E.

1983-03-01

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through cladmore » melting at 1370/sup 0/C.« less

Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending

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

Wang, Hong; Wang, Jy-An John

The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bendingmore » Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.« less

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL

DOEpatents

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

1959-01-27

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

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.

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.

Mechanistic Considerations Used in the Development of the PROFIT PCI Failure Model

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

Pankaskie, P. J.

A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) Interactions (PC!) failure model for estimating the probability of failure in !ransient increases in power (PROFIT) was developed. PROFIT is based on 1) standard statistical methods applied to available PC! fuel failure data and 2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that parameters in addition to power, transient increase in power, and burnup are needed to define PCI fuel failures in terms of probability estimates with known confidence limits. The PROFIT model, therefore, introduces an environmentalmore » and strain-rate dependent strain energy absorption to failure (SEAF) concept to account for the stress versus strain anomalies attributable to interstitial-disloction interaction effects in the Zircaloy cladding. Assuming that the power ramping rate is the operating corollary of strain-rate in the Zircaloy cladding, then the variables of first order importance in the PCI fuel failure phenomenon are postulated to be: 1. pre-transient fuel rod power, P{sub I}, 2. transient increase in fuel rod power, {Delta}P, 3. fuel burnup, Bu, and 4. the constitutive material property of the Zircaloy cladding, SEAF.« less

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

Manufacture of thin-walled clad tubes by pressure welding of roll bonded sheets

NASA Astrophysics Data System (ADS)

Schmidt, Hans Christian; Grydin, Olexandr; Stolbchenko, Mykhailo; Homberg, Werner; Schaper, Mirko

2017-10-01

Clad tubes are commonly manufactured by fusion welding of roll bonded metal sheets or, mechanically, by hydroforming. In this work, a new approach towards the manufacture of thin-walled tubes with an outer diameter to wall thickness ratio of about 12 is investigated, involving the pressure welding of hot roll bonded aluminium-steel strips. By preparing non-welded edges during the roll bonding process, the strips can be zip-folded and (cold) pressure welded together. This process routine could be used to manufacture clad tubes in a continuous process. In order to investigate the process, sample tube sections with a wall thickness of 2.1 mm were manufactured by U-and O-bending from hot roll bonded aluminium-stainless steel strips. The forming and welding were carried out in a temperature range between RT and 400°C. It was found that, with the given geometry, a pressure weld is established at temperatures starting above 100°C. The tensile tests yield a maximum bond strength at 340°C. Micrograph images show a consistent weld of the aluminium layer over the whole tube section.

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.

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

Montgomery, Rose; Scaglione, John M; Bevard, Bruce Balkcom

The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encounteredmore » during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.« less

Method for producing titanium aluminide weld rod

DOEpatents

Hansen, Jeffrey S.; Turner, Paul C.; Argetsinger, Edward R.

1995-01-01

A process for producing titanium aluminide weld rod comprising: attaching one end of a metal tube to a vacuum line; placing a means between said vacuum line and a junction of the metal tube to prevent powder from entering the vacuum line; inducing a vacuum within the tube; placing a mixture of titanium and aluminum powder in the tube and employing means to impact the powder in the tube to a filled tube; heating the tube in the vacuum at a temperature sufficient to initiate a high-temperature synthesis (SHS) reaction between the titanium and aluminum; and lowering the temperature to ambient temperature to obtain a intermetallic titanium aluminide alloy weld rod.

Fabrication and characterization of a hybrid four-hole AsSe₂-As₂S₅ microstructured optical fiber with a large refractive index difference.

PubMed

Cheng, Tonglei; Kanou, Yasuhire; Deng, Dinghuan; Xue, Xiaojie; Matsumoto, Morio; Misumi, Takashi; Suzuki, Takenobu; Ohishi, Yasutake

2014-06-02

A hybrid four-hole AsSe2-As2S5 microstructured optical fiber (MOF) with a large refractive index difference is fabricated by the rod-in-tube drawing technique. The core and the cladding are made from the AsSe2 glass and As2S5 glass, respectively. The propagation loss is ~1.8 dB/m and the nonlinear coefficient is ~2.03 × 10(4) km(-1)W(-1) at 2000 nm. Raman scattering is observed in the normal dispersion regime when the fiber is pumped by a 2 μm mode-locked picosecond fiber laser. Additionally, soliton is generated in the anomalous dispersion regime when the fiber is pumped by an optical parametric oscillator (OPO) at the pump wavelength of ~3000 nm.

Protected Nuclear Fuel Element

DOEpatents

Kittel, J. H.; Schumar, J. F.

1962-12-01

A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

Stress corrosion crack initiation of Zircaloy-4 cladding tubes in an iodine vapor environment during creep, relaxation, and constant strain rate tests

NASA Astrophysics Data System (ADS)

Jezequel, T.; Auzoux, Q.; Le Boulch, D.; Bono, M.; Andrieu, E.; Blanc, C.; Chabretou, V.; Mozzani, N.; Rautenberg, M.

2018-02-01

During accidental power transient conditions with Pellet Cladding Interaction (PCI), the synergistic effect of the stress and strain imposed on the cladding by thermal expansion of the fuel, and corrosion by iodine released as a fission product, may lead to cladding failure by Stress Corrosion Cracking (SCC). In this study, internal pressure tests were conducted on unirradiated cold-worked stress-relieved Zircaloy-4 cladding tubes in an iodine vapor environment. The goal was to investigate the influence of loading type (constant pressure tests, constant circumferential strain rate tests, or constant circumferential strain tests) and test temperature (320, 350, or 380 °C) on iodine-induced stress corrosion cracking (I-SCC). The experimental results obtained with different loading types were consistent with each other. The apparent threshold hoop stress for I-SCC was found to be independent of the test temperature. SEM micrographs of the tested samples showed many pits distributed over the inner surface, which tended to coalesce into large pits in which a microcrack could initiate. A model for the time-to-failure of a cladding tube was developed using finite element simulations of the viscoplastic mechanical behavior of the material and a modified Kachanov's damage growth model. The times-to-failure predicted by this model are consistent with the experimental data.

PLUTONIUM FUEL RODS FOR PREPARATION OF TRANSPLUTONIC ELEMENTS

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

Bailey, W.J.

1962-02-01

Production by coextrusion of metallurgically bonded, Alclad, Al-7.35 wt% Pu alloy fuel rods with integral ends is discussed. The rods had a diameter of 0.94 in., length of, 60 in., and a nominal cladding thickness of 0.070 in. The Pu concentration was maintained at 83.3 g/rod. The coextrusion billets can be assembled with fuel cores in the as-cast condition. The casting hot-tops can be returned to the process stream. The process is useful for preparing transplutonic elements and production of high-exposure Pu. (J.R.D.)

ANALYSIS AND EXAMINATION OF MOX FUEL FROM NONPROLIFERATION PROGRAMS

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

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

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

Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

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

Billone, M. C.; Burtseva, T. A.

2016-08-30

The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

FRAPCON analysis of cladding performance during dry storage operations

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

Richmond, David J.; Geelhood, Kenneth J.

There is an increasing need in the U.S. and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations (ISFSI) or interim storage sites. The NRC limits cladding temperature to 400°C while maintaining cladding hoop stress below 90 MPa in an effort to avoid radial hydride reorientation. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400 °C. Results were representative of the majority of U.S. LWR fuel. They conservativelymore » showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.« less

Residual Stress Measurement and the Effect of Heat Treatment in Cladded Control Rod Drive Specimens

NASA Astrophysics Data System (ADS)

Bowman, Ashley; Kingston, Ed; Katsuyama, Jinya; Udagawa, Makoto; Onizawa, Kunio

This paper presents results of residual stress measurements and modelling within the cladding and J-groove weld of Control Rod Drive (CRD) specimens in the as-welded and Post Weld Heat Treated (PWHT) states. Knowledge of the residual stresses present in CRD nozzles is critical when modelling the fracture mechanics of failures of nuclear power plant components to dictate inspections intervals and optimise plant downtime. The specimens comprised of ferritic steel blocks with 309L stainless steel cladding and a single J-groove weld attaching the 304 stainless steel nozzles. Multiple measurements were made through the thickness of the specimens in order to give biaxial residual stress profiles through all the different fusion boundaries. The results show the effect of PWHT in reducing residual stresses both in the weld and ferritic material. The beneficial use of measurements is highlighted to provide confidence in the modelled results and prevent over conservatism in integrity calculations, costing unnecessary time and money.

Validating the BISON fuel performance code to integral LWR experiments

DOE PAGES

Williamson, R. L.; Gamble, K. A.; Perez, D. M.; ...

2016-03-24

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to datemore » for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Design of a fuel element for a lead-cooled fast reactor

NASA Astrophysics Data System (ADS)

Sobolev, V.; Malambu, E.; Abderrahim, H. Aït

2009-03-01

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg-1 of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg-1 of HM.

Risk Assessment of Structural Integrity of Transportation Casks after Extended Storage

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

Ibarra, Luis; Medina, Ricardo; Yang, Haori

This study assessed the risk of loss of structural integrity of transportation casks and fuel cladding after extended storage. Although it is known that fuel rods discharged from NPPs have a small percentage of rod cladding defects, the behavior of fuel cladding and the structural elements of assemblies during transportation after long-term storage is not well understood. If the fuel degrades during extended storage, it could be susceptible to damage from vibration and impact loads during transport operations, releasing fission-product gases into the canister or the cask interior (NWTRB 2010). Degradation of cladding may occur due to mechanisms associated withmore » hydrogen embrittlement, delayed hydride cracking, low temperature creep, and stress corrosion cracking (SCC) that may affect fuel cladding and canister components after extended storage of hundreds of years. Over extended periods at low temperatures, these mechanisms affect the ductility, strength, and fracture toughness of the fuel cladding, which becomes brittle. For transportation purposes, the fuel may be transferred from storage to shipping casks, or dual-purpose casks may be used for storage and transportation. Currently, most of the transportation casks will be the former case. A risk assessment evaluation is conducted based on results from experimental tests and simulations with advanced numerical models. A novel contribution of this study is the evaluation of the combined effect of component aging and vibration/impact loads in transportation scenarios. The expected levels of deterioration will be obtained from previous and current studies on the effect of aging on fuel and cask components. The emphasis of the study is placed on the structural integrity of fuel cladding and canisters.« less

Burst Ductility of Zirconium Clads: The Defining Role of Residual Stress

NASA Astrophysics Data System (ADS)

Kumar, Gulshan; Kanjarla, A. K.; Lodh, Arijit; Singh, Jaiveer; Singh, Ramesh; Srivastava, D.; Dey, G. K.; Saibaba, N.; Doherty, R. D.; Samajdar, Indradev

2016-08-01

Closed end burst tests, using room temperature water as pressurizing medium, were performed on a number of industrially produced zirconium (Zr) clads. A total of 31 samples were selected based on observed differences in burst ductility. The latter was represented as total circumferential elongation or TCE. The selected samples, with a range of TCE values (5 to 35 pct), did not show any correlation with mechanical properties along axial direction, microstructural parameters, crystallographic textures, and outer tube-surface normal ( σ 11) and shear ( τ 13) components of the residual stress matrix. TCEs, however, had a clear correlation with hydrostatic residual stress ( P h), as estimated from tri-axial stress analysis on the outer tube surface. Estimated P h also scaled with measured normal stress ( σ 33) at the tube cross section. An elastic-plastic finite element model with ductile damage failure criterion was developed to understand the burst mechanism of zirconium clads. Experimentally measured P h gradients were imposed on a solid element continuum finite element (FE) simulation to mimic the residual stresses present prior to pressurization. Trends in experimental TCEs were also brought out with computationally efficient shell element-based FE simulations imposing the outer tube-surface P h values. Suitable components of the residual stress matrix thus determined the burst performance of the Zr clads.

Nuclear reactor safety device

DOEpatents

Hutter, E.

1983-08-15

A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

Determination of He and D permeability of neutron-irradiated SiC tubes to examine the potential for release due to micro-cracking

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

Katoh, Yutai; Hu, Xunxiang; Koyanagi, Takaaki

Driven by the need to enlarge the safety margins of light water reactors in both design-basis and beyond-design-basis accident scenarios, the research and development of accident-tolerant fuel (ATF) has become an importance topic in the nuclear engineering and materials community. Continuous SiC fiber-reinforced SiC matrix ceramic composites are under consideration as a replacement for traditional zirconium alloy cladding owing to their high-temperature stability, chemical inertness, and exceptional irradiation resistance. Among the key technical feasibility issues, potential failure of the fission product containment due to probabilistic penetrating cracking has been identified as one of the two most critical feasibility issues, togethermore » with the radiolysisassisted hydrothermal corrosion of SiC. The experimental capability to evaluate the hermeticity of SiC-based claddings is an urgent need. In this report, we present the development of a comprehensive permeation testing station established in the Low Activation Materials Development and Analysis laboratory at Oak Ridge National Laboratory. Preliminary results for the hermeticity evaluation of un-irradiated monolithic SiC tubes, uncoated and coated SiC/SiC composite tubes, and neutron-irradiated monolithic SiC tubes at room temperature are exhibited. The results indicate that this new permeation testing station is capable of evaluating the hermeticity of SiC-based tubes by determining the helium and deuterium permeation flux as a function of gas pressure at a high resolution of 8.07 x 10 -12 atm-cc/s for helium and 2.83 x 10 -12 atm-cc/s for deuterium, respectively. The detection limit of this system is sufficient to evaluate the maximum allowable helium leakage rate of lab-scale tubular samples, which is linearly extrapolated from the evaluation standard used for a commercial as-manufactured light water reactor fuel rod at room temperature. The un-irradiated monolithic SiC tube is hermetic, as is manifested by the un-detectable deuterium permeation flux at various feeding gas pressures. A large helium leakage rate was detected for the uncoated SiC/SiC composite tube exposed to atmosphere, indicating it is inherently not hermetic. The hermeticity of coated SiC/SiC composite tubes is strongly dependent on the coating materials and the preparation of the substrate SiC/SiC composite samples. To simulate the practical application environment, monolithic CVD SiC tubes were exposed to neutron irradiation at the High Flux Isotope Reactor under high heat flux from the internal surface to the external surface. Although finite element analysis and resonant ultrasound spectroscopy measurement indicated that the combined neutron irradiation and high heat flux gave rise to a high probability of cracking within the sample, the hermeticity evaluation of the tested sample still exhibited gas tightness, emphasizing that SiC cracking is inherently a statistical phenomenon. The developed permeation testing station is capable of measuring the gas permeation flux in the range of interest with full confidence based on the presented results. It is considered a critical pre- /post-irradiation examination technique to characterize SiC-based cladding materials in asreceived and irradiated states to aid the research and development of ATF.« less

Effects of pretreatment processes for Zr electrorefining of oxidized Zircaloy-4 cladding tubes

NASA Astrophysics Data System (ADS)

Hwa Lee, Chang; Lee, Yoo Lee; Jeon, Min Ku; Choi, Yong Taek; Kang, Kweon Ho; Park, Geun Il

2014-06-01

The effect of pretreatment processes for the Zr electrorefining of oxidized Zircaloy-4 cladding tubes is examined in LiCl-KCl-ZrCl4 molten salts at 500 °C. The cyclic voltammetries reveal that the Zr dissolution kinetics is highly dependent on the thickness of a Zr oxide layer formed at 500 °C under air atmosphere. For the Zircaloy-4 tube covered with a 1 μm thick oxide layer, the Zr dissolution process is initiated from a non-stoichiometric Zr oxide surface through salt treatment at an open circuit potential in the molten salt electrolyte. The Zr dissolution of the samples in the middle range of oxide layer thickness appears to be more effectively derived by the salt treatment coupled with an anodic potential application at an oxidation potential of Zr. A modification of the process scheme offers an applicability of Zr electrorefining for the treatment of oxidized cladding hull wastes.

Nuclear reactor fuel element with vanadium getter on cladding

DOEpatents

Johnson, Carl E.; Carroll, Kenneth G.

1977-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

Development and Validation of Accident Models for FeCrAl Cladding

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

Gamble, Kyle Allan Lawrence; Hales, Jason Dean

2016-08-01

The purpose of this milestone report is to present the work completed in regards to material model development for FeCrAl cladding and highlight the results of applying these models to Loss of Coolant Accidents (LOCA) and Station Blackouts (SBO). With the limited experimental data available (essentially only the data used to create the models) true validation is not possible. In the absence of another alternative, qualitative comparisons during postulated accident scenarios between FeCrAl and Zircaloy-4 cladded rods have been completed demonstrating the superior performance of FeCrAl.

Development of advanced strain diagnostic techniques for reactor environments.

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

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.

2013-02-01

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding.more » During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.« less

Nuclear reactor control apparatus

DOEpatents

Sridhar, Bettadapur N.

1983-11-01

Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

DOE PAGES

Gamble, Kyle A.; Barani, Tommaso; Pizzocri, David; ...

2017-04-30

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

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

Gamble, Kyle A.; Barani, Tommaso; Pizzocri, David

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less

Investigation of a quadrupole ultra-high vacuum ion pump

NASA Technical Reports Server (NTRS)

Schwarz, H. J.

1974-01-01

The new nonmagnetic ion pump resembles the quadrupole ionization gage. The dimensions are larger, and hyperbolically shaped electrodes replace the four rods. Their surfaces follow y sq. = 36 + x sq. (x, y in centimeters). The electrodes, 55 cm long, are positioned lengthwise in a tube. At one end a cathode emits electrons; at the other end a narrowly wound flat spiral of tungsten clad with titanium on cathode potential can be heated for titanium evaporation. Electrons accelerated by a dc potential of the surface electrodes oscillate between the ends on rotational trajectories, if a high frequency potential superimposed on the dc potential is properly adjusted. Pumping speeds (4-100 liter/sec) for different gases at different peak voltages (1000-3000V) at corresponding frequencies (57-100 MHz), and at different pressures 0.00001 to the minus 9 power Torr were observed. The lowest pressure reached was below 10 to the minus 10 power Torr.

Broadband mid-infrared supercontinuum generation in novel As2Se3-As2Se2 S step-index fibers

NASA Astrophysics Data System (ADS)

Wang, Yingying; Dai, Shixun; Han, Xin; Zhang, Peiqing; Liu, Yongxing; Wang, Xunsi; Sun, Shaochao

2018-03-01

We experimentally demonstrate the mid-infrared supercontinuum generation in a chalcogenide step-index fiber consisting of an As2Se3 core and an As2Se2 S cladding. The fiber with the core diameter of 21 μm was fabricated through the rod-in-tube technique and fiber-drawing process. The effect of pump wavelength, fiber length, and pump power on the spectral bandwidth and output power of the supercontinuum spectra generated from the fiber pumped by the ultrashort pulses of ∼ 150 fs with a repetition rate of 1000 Hz was systematically investigated. When pumping a 12-cm-long fiber at a wavelength of 6 . 5 μm with 14 mW pump laser power, a broadband supercontinuum spanning from 2 . 0 μm to 12 . 7 μm with an output power of 300 μW was obtained.

Clip gage attachment for frictionless measurement of displacement during high-temperature mechanical testing

DOEpatents

Alexander, David J.

1994-01-01

An attachment for placement between a test specimen and a remote clip gage extensometer providing improved fracture toughness tests of materials at elevated temperature. Using a cylindrical tube and axial rod in new relationship, the device transfers the displacement signal of the fracture toughness test specimen directly to a clip gage extensometer located outside the high temperature furnace. Virtually frictionless operation is assured by having the test specimen center one end of the rod in one end of the tube, while the clip gage extensometer arms center the other end of the rod in the other end of the tube. By providing positive control over both ends of both rod and tube, the attachment may be operated in orientations other than vertical.

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

DOE PAGES

Cinbiz, M. Nedim; Brown, Nicholas R.; Terrani, Kurt A.; ...

2017-06-03

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failuremore » greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.« less

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

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

Cinbiz, M. Nedim; Brown, Nicholas R.; Terrani, Kurt A.

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failuremore » greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.« less

Integrated head package for top mounted nuclear instrumentation

DOEpatents

Malandra, Louis J.; Hornak, Leonard P.; Meuschke, Robert E.

1993-01-01

A nuclear reactor such as a pressurized water reactor has an integrated head package providing structural support and increasing shielding leading toward the vessel head. A reactor vessel head engages the reactor vessel, and a control rod guide mechanism over the vessel head raises and lowers control rods in certain of the thimble tubes, traversing penetrations in the reactor vessel head, and being coupled to the control rods. An instrumentation tube structure includes instrumentation tubes with sensors movable into certain thimble tubes disposed in the fuel assemblies. Couplings for the sensors also traverse penetrations in the reactor vessel head. A shroud is attached over the reactor vessel head and encloses the control rod guide mechanism and at least a portion of the instrumentation tubes when retracted. The shroud forms a structural element of sufficient strength to support the vessel head, the control rod guide mechanism and the instrumentation tube structure, and includes radiation shielding material for limiting passage of radiation from retracted instrumentation tubes. The shroud is thicker at the bottom adjacent the vessel head, where the more irradiated lower ends of retracted sensors reside. The vessel head, shroud and contents thus can be removed from the reactor as a unit and rested safely and securely on a support.

Compatibility of refractory materials for nuclear reactor poison control systems

NASA Technical Reports Server (NTRS)

Sinclair, J. H.

1974-01-01

Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

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

Gamble, Kyle Allan Lawrence; Novascone, Stephen Rhead; Gardner, Russell James

2016-06-01

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation andmore » power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Tube manufacturing and characterization of oxide dispersion strengthened ferritic steels

NASA Astrophysics Data System (ADS)

Ukai, Shigeharu; Mizuta, Shunji; Yoshitake, Tunemitsu; Okuda, Takanari; Fujiwara, Masayuki; Hagi, Shigeki; Kobayashi, Toshimi

2000-12-01

Oxide dispersion strengthened (ODS) ferritic steels have an advantage in radiation resistance and superior creep rupture strength at elevated temperature due to finely distributed Y2O3 particles in the ferritic matrix. Using a basic composition of low activation ferritic steel (Fe-12Cr-2W-0.05C), cladding tube manufacturing by means of pilger mill rolling and subsequent recrystallization heat-treatment was conducted while varying titanium and yttria contents. The recrystallization heat-treatment, to soften the tubes hardened due to cold-rolling and to subsequently improve the degraded mechanical properties, was demonstrated to be effective in the course of tube manufacturing. For a titanium content of 0.3 wt% and yttria of 0.25 wt%, improvement of the creep rupture strength can be attained for the manufactured cladding tubes. The ductility is also adequately maintained.

Part Repairing Using A Hybrid Manufacturing System (Preprint)

DTIC Science & Technology

2007-03-01

laser . The laser processing parameters for cladding steel H13 powder were 600W with a stand-off distance from the nozzle to the top of the clad of 0.5...Journal of Materials Processing Technology, 2002:122, 63-68. [11]Richter, K., Orban, S., and Nowotny, S., Laser cladding of the titanium alloy TI6242...was used to repair the corroded steam generator tubes in nuclear plants [9], and turbine blades were repaired using the laser cladding process [10

Clip gage attachment for frictionless measurement of displacement during high-temperature mechanical testing

DOEpatents

Alexander, D.J.

1994-01-04

An attachment for placement between a test specimen and a remote clip gage extensometer providing improved fracture toughness tests of materials at elevated temperature is described. Using a cylindrical tube and axial rod in new relationship, the device transfers the displacement signal of the fracture toughness test specimen directly to a clip gage extensometer located outside the high temperature furnace. Virtually frictionless operation is assured by having the test specimen center one end of the rod in one end of the tube, while the clip gage extensometer arms center the other end of the rod in the other end of the tube. By providing positive control over both ends of both rod and tube, the attachment may be operated in orientations other than vertical. 1 figure.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-01

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

On the control of the crystallographic texture in cladding tubes from Zr-based alloys for nuclear reactor

NASA Astrophysics Data System (ADS)

Isaenkova, M.; Perlovich, Yu.; Fesenko, V.

2016-10-01

This paper summarizes researches of authors, directed to the development of the methodological basis of X-ray studies as applied to zirconium alloys and on the systematization of new experimental results obtained using developed methods. The paper describes regularities of crystallographic texture formation in cladding tubes from zirconium alloys and their substructure inhomogeneity at various stages of manufacture, i.e. at hot and cold deformation, recrystallization, phase transformations and interaction of the above processes. The special attention is payed to possibilities of control the crystallographic texture of tubes at successive stages of their technological treatment.

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

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

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

1994-12-31

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

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

Lee, Young-Ho; Byun, Thak Sang

Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate themore » tribological compatibilities of selfmated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.« less

Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

2018-01-01

Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been necessary to perform a careful design study of the probe geometry. For this, finite element analysis (FEA) has been performed in combination with practical validation tests on representative fuel dummies with machined flaws to find the probe geometry that best detects a hidden flaw. Tests performed thus far show that gaps down to 25 μm thickness can be detected with good repeatability and good discrimination from lift-off signals.

Fuel Performance Calculations for FeCrAl Cladding in BWRs

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

George, Nathan; Sweet, Ryan; Maldonado, G. Ivan

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behaviormore » of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.« less

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients

NASA Astrophysics Data System (ADS)

Gallais, L.; Burla, R.; Martin, F.; Richaud, J. C.; Volle, G.; Pontillon, M.; Capdevila, H.; Pontillon, Y.

2018-01-01

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients.

PubMed

Gallais, L; Burla, R; Martin, F; Richaud, J C; Volle, G; Pontillon, M; Capdevila, H; Pontillon, Y

2018-01-01

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

Low-loss single-mode hollow-core fiber with anisotropic anti-resonant elements.

PubMed

Habib, Md Selim; Bang, Ole; Bache, Morten

2016-04-18

A hollow-core fiber using anisotropic anti-resonant tubes in the cladding is proposed for low loss and effectively single-mode guidance. We show that the loss performance and higher-order mode suppression is significantly improved by using symmetrically distributed anisotropic anti-resonant tubes in the cladding, elongated in the radial direction, when compared to using isotropic, i.e. circular, anti-resonant tubes. The effective single-mode guidance of the proposed fiber is achieved by enhancing the coupling between the cladding modes and higher-order-core modes by suitably engineering the anisotropic anti-resonant elements. With a silica-based fiber design aimed at 1.06 µm, we show that the loss extinction ratio between the higher-order core modes and the fundamental core mode can be more than 1000 in the range 1.0-1.65 µm, while the leakage loss of the fundamental core mode is below 15 dB/km in the same range.

An elasto-plastic fracture mechanics based model for assessment of hydride embrittlement in zircaloy cladding tubes

NASA Astrophysics Data System (ADS)

Nilsson, Karl-Fredrik; Jakšić, Nikola; Vokál, Vratko

2010-01-01

This paper describes a finite element based fracture mechanics model to assess how hydrides affect the integrity of zircaloy cladding tubes. The hydrides are assumed to fracture at a low load whereas the propagation of the fractured hydrides in the matrix material and failure of the tube is controlled by non-linear fracture mechanics and plastic collapse of the ligaments between the hydrides. The paper quantifies the relative importance of hydride geometrical parameters such as size, orientation and location of individual hydrides and interaction between adjacent hydrides. The paper also presents analyses for some different and representative multi-hydride configurations. The model is adaptable to general and complex crack configurations and can therefore be used to assess realistic hydride configurations. The mechanism of cladding failure is by plastic collapse of ligaments between interacting fractured hydrides. The results show that the integrity can be drastically reduced when several radial hydrides form continuous patterns.

Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

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

Montgomery, Robert O.; Capps, Nathan A.; Sunderland, Dion J.

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code thatmore » is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

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

Jason Hales; Various

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale codemore » that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

CONTROL ROD

DOEpatents

Zinn, W.H.; Ross, H.V.

1958-11-18

A control rod is described for a nuclear reactor. In certaln reactor designs it becomes desirable to use a control rod having great width but relatively llttle thickness. This patent is addressed to such a need. The neutron absorbing material is inserted in a triangular tube, leaving volds between the circular insert and the corners of the triangular tube. The material is positioned within the tube by the use of dummy spacers to achleve the desired absorption pattern, then the ends of the tubes are sealed with suitable plugs. The tubes may be welded or soldered together to form two flat surfaces of any desired width, and covered with sheetmetal to protect the tubes from damage. This design provides a control member that will not distort under the action of outside forces or be ruptured by gases generated within the jacketed control member.

Optimized manufacture of nuclear fuel cladding tubes by FEA of hot extrusion and cold pilgering processes

NASA Astrophysics Data System (ADS)

Gaillac, Alexis; Ly, Céline

2018-05-01

Within the forming route of Zirconium alloy cladding tubes, hot extrusion is used to deform the forged billets into tube hollows, which are then cold rolled to produce the final tubes with the suitable properties for in-reactor use. The hot extrusion goals are to give the appropriate geometry for cold pilgering, without creating surface defects and microstructural heterogeneities which are detrimental for subsequent rolling. In order to ensure a good quality of the tube hollows, hot extrusion parameters have to be carefully chosen. For this purpose, finite element models are used in addition to experimental tests. These models can take into account the thermo-mechanical coupling conditions obtained in the tube and the tools during extrusion, and provide a good prediction of the extrusion load and the thermo-mechanical history of the extruded product. This last result can be used to calculate the fragmentation of the microstructure in the die and the meta-dynamic recrystallization after extrusion. To further optimize the manufacturing route, a numerical model of the cold pilgering process is also applied, taking into account the complex geometry of the tools and the pseudo-steady state rolling sequence of this incremental forming process. The strain and stress history of the tube during rolling can then be used to assess the damage risk thanks to the use of ductile damage models. Once validated vs. experimental data, both numerical models were used to optimize the manufacturing route and the quality of zirconium cladding tubes. This goal was achieved by selecting hot extrusion parameters giving better recrystallized microstructure that improves the subsequent formability. Cold pilgering parameters were also optimized in order to reduce the potential ductile damage in the cold rolled tubes.

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

Process for producing clad superconductive materials

DOEpatents

Cass, Richard B.; Ott, Kevin C.; Peterson, Dean E.

1992-01-01

A process for fabricating superconducting composite wire by the steps of placing a superconductive precursor admixture capable of undergoing a self propagating combustion in stoichiometric amounts sufficient to form a superconductive product within a metal tube, sealing one end of said tube, igniting said superconductive precursor admixture whereby said superconductive precursor admixture endburns along the length of the admixture, and cross-section reducing said tube at a rate substantially equal to the rate of burning of said superconductive precursor admixture and at a point substantially planar with the burnfront of the superconductive precursor mixture, whereby a clad superconductive product is formed in situ, the product characterized as superconductive without a subsequent sintering stage, is disclosed.

Chemical Dosimeter Tube With Coaxial Sensing Rod

NASA Technical Reports Server (NTRS)

Lueck, Dale E.

1993-01-01

Improved length-of-stain (LOS) chemical dosimeter indicates total dose of chemical vapor in air. Made with rods and tubes of various diameters to obtain various sensitivities and dynamic ranges. Sensitivity larger and dose range smaller when more room for diffusion in gap between tube and rod. Offers greater resistance to changing of color of exposed dye back to color of unexposed condition, greater sensitivity, and higher degree of repeatability. Developed to measure doses of gaseous HCI, dosimeter modified by use of other dyes to indicate doses of other chemical vapors.

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

3D-Printed Broadband Dielectric Tube Terahertz Waveguide with Anti-Reflection Structure

NASA Astrophysics Data System (ADS)

Vogt, Dominik Walter; Leonhardt, Rainer

2016-11-01

We demonstrate broadband, low loss, and close-to-zero dispersion guidance of terahertz (THz) radiation in a dielectric tube with an anti-reflection structure (AR-tube waveguide) in the frequency range from 0.2 to 1.0 THz. The anti-reflection structure (ARS) consists of close-packed cones in a hexagonal lattice arranged on the outer surface of the tube cladding. The feature size of the ARS is in the order of the wavelength between 0.2 and 1.0 THz. The waveguides are fabricated with the versatile and cost efficient 3D-printing method. Terahertz time-domain spectroscopy (THz-TDS) measurements as well as 3D finite-difference time-domain simulations (FDTD) are performed to extensively characterize the AR-tube waveguides. Spectrograms, attenuation spectra, effective phase refractive indices, and the group-velocity dispersion parameters β 2 of the AR-tube waveguides are presented. Both the experimental and numerical results confirm the extended bandwidth and smaller group-velocity dispersion of the AR-tube waveguide compared to a low loss plain dielectric tube THz waveguide. The AR-tube waveguide prototypes show an attenuation spectrum close to the theoretical limit given by the infinite cladding tube waveguide.

Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

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

Patra, Anirban; Tome, Carlos

This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

Numerical Modeling of Tube Forming by HPTR Cold Pilgering Process

NASA Astrophysics Data System (ADS)

Sornin, D.; Pachón-Rodríguez, E. A.; Vanegas-Márquez, E.; Mocellin, K.; Logé, R.

2016-09-01

For new fast-neutron sodium-cooled Generation IV nuclear reactors, the candidate cladding materials for the very strong burn-up are ferritic and martensitic oxide dispersion strengthened grades. Classically, the cladding tube is cold formed by a sequence of cold pilger milling passes with intermediate heat treatments. This process acts upon the geometry and the microstructure of the tubes. Consequently, crystallographic texture, grain sizes and morphologies, and tube integrity are highly dependent on the pilgering parameters. In order to optimize the resulting mechanical properties of cold-rolled cladding tubes, it is essential to have a thorough understanding of the pilgering process. Finite Element Method (FEM) models are used for the numerical predictions of this task; however, the accuracy of the numerical predictions depends not only on the type of constitutive laws but also on the quality of the material parameters identification. Therefore, a Chaboche-type law which parameters have been identified on experimental observation of the mechanical behavior of the material is used here. As a complete three-dimensional FEM mechanical analysis of the high-precision tube rolling (HPTR) cold pilgering of tubes could be very expensive, only the evolution of geometry and deformation is addressed in this work. The computed geometry is compared to the experimental one. It is shown that the evolution of the geometry and deformation is not homogeneous over the circumference. Moreover, it is exposed that the strain is nonhomogeneous in the radial, tangential, and axial directions. Finally, it is seen that the dominant deformation mode of a material point evolves during HPTR cold pilgering forming.

Active Tube-Shaped Actuator with Embedded Square Rod-Shaped Ionic Polymer-Metal Composites for Robotic-Assisted Manipulation

PubMed Central

Liu, Jiayu; Zhu, Denglin; Chen, Hualing

2018-01-01

This paper reports a new technique involving the design, fabrication, and characterization of an ionic polymer-metal composite- (IPMC-) embedded active tube, which can achieve multidegree-of-freedom (MODF) bending motions desirable in many applications, such as a manipulator and an active catheter. However, traditional strip-type IPMC actuators are limited in only being able to generate 1-dimensional bending motion. So, in this paper, we try to develop an approach which involves molding or integrating rod-shaped IPMC actuators into a soft silicone rubber structure to create an active tube. We modified the Nafion solution casting method and developed a complete sequence of a fabrication process for rod-shaped IPMCs with square cross sections and four insulated electrodes on the surface. The silicone gel was cured at a suitable temperature to form a flexible tube using molds fabricated by 3D printing technology. By applying differential voltages to the four electrodes of each IPMC rod-shaped actuator, MDOF bending motions of the active tube can be generated. Experimental results show that such IPMC-embedded tube designs can be used for developing robotic-assisted manipulation. PMID:29770160

Active Tube-Shaped Actuator with Embedded Square Rod-Shaped Ionic Polymer-Metal Composites for Robotic-Assisted Manipulation.

PubMed

Wang, Yanjie; Liu, Jiayu; Zhu, Denglin; Chen, Hualing

2018-01-01

This paper reports a new technique involving the design, fabrication, and characterization of an ionic polymer-metal composite- (IPMC-) embedded active tube, which can achieve multidegree-of-freedom (MODF) bending motions desirable in many applications, such as a manipulator and an active catheter. However, traditional strip-type IPMC actuators are limited in only being able to generate 1-dimensional bending motion. So, in this paper, we try to develop an approach which involves molding or integrating rod-shaped IPMC actuators into a soft silicone rubber structure to create an active tube. We modified the Nafion solution casting method and developed a complete sequence of a fabrication process for rod-shaped IPMCs with square cross sections and four insulated electrodes on the surface. The silicone gel was cured at a suitable temperature to form a flexible tube using molds fabricated by 3D printing technology. By applying differential voltages to the four electrodes of each IPMC rod-shaped actuator, MDOF bending motions of the active tube can be generated. Experimental results show that such IPMC-embedded tube designs can be used for developing robotic-assisted manipulation.

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

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

R. L. Williamson; D. A. Knoll

2009-09-01

A powerful multidimensional fuels performance capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth , gap heat transfer, and gap/plenum gas behavior during irradiation. The various modeling capabilities are demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multi-pellet fuel rod, during both steady and transient operation. Computational results demonstrate the importancemore » of a multidimensional fully-coupled thermomechanics treatment. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermo-mechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.« less

Microstructural characterization of AA5183 aluminum clad AISI 1018 steel prepared by electro spark deposition

NASA Astrophysics Data System (ADS)

Rastkerdar, E.; Aghajani, H.; Kianvash, A.; Sorrell, C. C.

2018-04-01

The application of a simple and effective technique, electro spark deposition (ESD), to create aluminum clad steel plate has been studied. AA5183 aluminum rods were used as the rotating electrode for cladding of the AISI 1018 steel. The microstructure of the interfacial zone including the intermetallic compounds (IMC) layer and the clad metal have been investigated by scanning electron microscopy (SEM) equipped with energy dispersive spectroscopy (EDS) and transmission electron microscopy (TEM and STEM). According to the results sound aluminum clad with thickness up to 25–30 μm can be achieved. Very thin (<4 μm) IMC layer was formed at the Al/Fe interface and the structural (electron diffraction pattern) and chemical analysis (STEM) conducted by TEM confirmed that the layer is constituted of Fe rich phases, both implying a much improved mechanical properties. Investigation of the orientations of phases at the interfacial zone confirmed absence of any preferred orientation.

Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

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

Asmolov, V.; Yegorova, L.; Kaplar, E.

1998-03-01

Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in thismore » paper.« less

RIA simulation tests using driver tube for ATF cladding

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

Cinbiz, Mahmut N.; Brown, N. R.; Lowden, R. R.

Pellet-cladding mechanical interaction (PCMI) is a potential failure mechanism for accident-tolerant fuel (ATF) cladding candidates during a reactivity-initiated accident (RIA). This report summarizes Fiscal Year (FY) 2017 research activities that were undertaken to evaluate the PCMI-like hoop-strain-driven mechanical response of ATF cladding candidates. To achieve various RIA-like conditions, a modified-burst test (MBT) device was developed to produce different mechanical pulses. The calibration of the MBT instrument was accomplished by performing mechanical tests on unirradiated Generation-I iron-chromium-aluminum (FeCrAl) alloy samples. Shakedown tests were also conducted in both FY 2016 and FY 2017 using unirradiated hydrided ZIRLO™ tube samples. This milestone reportmore » focuses on testing of ATF materials, but the benchmark tests with hydrided ZIRLO™ tube samples are documented in a recent journal article.a For the calibration and benchmark tests, the hoop strain was monitored using strain gauges attached to the sample surface in the hoop direction. A novel digital image correlation (DIC) system composed of a single high-speed camera and an array of six mirrors was developed for the MBT instrument to better resolve the failure behavior of samples and to provide useful data for validation of high-fidelity modeling and simulation tools. The DIC system enable a 360° view of a sample’s outer surface. This feature was added to the instrument to determine the precise failure location on a sample’s surface for strain predictions. The DIC system was tested on several silicon carbide fiber/silicon carbide matrix (SiC/SiC) composite tube samples at various pressurization rates of the driver tube (which correspond to the strain rates for the samples). The hoop strains for various loading conditions were determined for the SiC/SiC composite tube samples. Future work is planned to enhance understanding of the failure behavior of the ATF cladding candidates of age-hardened FeCrAl alloys and SiC/SiC composites in detail during RIA conditions informed by the computational studies performed under the US Department of Energy Office of Nuclear Energy Advanced Fuels Campaign. The testing instrument and the new DIC system will be further developed to reach different stress-state conditions and to perform tests at elevated temperatures.« less

Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux

NASA Astrophysics Data System (ADS)

Petrie, Christian M.; Koyanagi, Takaaki; McDuffee, Joel L.; Deck, Christian P.; Katoh, Yutai; Terrani, Kurt A.

2017-08-01

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300-350 °C under a representative high heat flux (∼0.66 MW/m2) during one cycle of irradiation in an un-instrumented ;rabbit; capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb the expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. The success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.

Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux

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

Petrie, Christian M.; Koyanagi, Takaaki; McDuffee, Joel L.

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300–350 °C under a representative high heat flux (~0.66 MW/m 2) during one cycle of irradiation in an un-instrumented “rabbit” capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb themore » expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. Furthermore, the success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.« less

Experimental and Numerical Investigation of Pressure Drop in Silicon Carbide Fuel Rod for Application in Pressurized Water Reactors

NASA Astrophysics Data System (ADS)

Abir, Ahmed Musafi

Spacer grids are used in Pressurized Water Reactors (PWRs) fuel assemblies which enhances heat transfer from fuel rods. However, there remain regions of low turbulence in between the spacer grids. To enhance turbulence in these regions surface roughness is applied on the fuel rod walls. Meyer [1] used empirical correlations to predict heat transfer and friction factor for artificially roughened fuel rod bundles at High Performance Light Water Reactors (LWRs). Their applicability was tested by Carrilho at University of South Carolina's (USC) Single Heated Element Loop Tester (SHELT). He attained a heat transfer and friction factor enhancement of 50% and 45% respectively, using Inconel nuclear fuel rods with square transverse ribbed surface. Following him Najeeb conducted a similar study due to three dimensional diamond shaped blocks in turbulent flow. He recorded a maximum heat transfer enhancement of 83%. At present, several types of materials are being used for fuel rod cladding including Zircaloy, Uranium oxide, etc. But researchers are actively searching for new material that can be a more practical alternative. Silicon Carbide (SiC) has been identified as a material of interest for application as fuel rod cladding [2]. The current study deals with the experimental investigation to find out the friction factor increase of a SiC fuel rod with 3D surface roughness. The SiC rod was tested at USC's SHELT loop. The experiment was conducted in turbulent flowing Deionized (DI) water at steady state conditions. Measurements of Flow rate and pressure drop were made. The experimental results were also validated by Computational Fluid Dynamics (CFD) analysis in ANSYS Fluent. To simplify the CFD analysis and to save computational resources the 3D roughness was approximated as a 2D one. The friction factor results of the CFD investigation was found to lie within +/-8% of the experimental results. A CFD model was also run with the energy equation turned on, and a heat generation of 8 kW applied to the rod. A maximum heat transfer enhancement of 18.4% was achieved at the highest flow rate investigated (i.e. Re=109204).

The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions

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

Cinbiz, Mahmut N; Brown, Nicholas R; Terrani, Kurt A

2017-01-01

This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding was subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings was compared to that of hydrided zirconium-based nuclear fuel cladding alloy. The hoop strain evolution during pressure pulses were collected in situ; the permanent diametral strains of bothmore » accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture.« less

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

Stimpson, Shane G; Powers, Jeffrey J; Clarno, Kevin T

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity, multiphysics simulations of light water reactors (LWRs) by coupling a variety of codes within the Virtual Environment for Reactor Analysis (VERA). One of the primary goals of CASL is to predict local cladding failure through pellet-clad interaction (PCI). This capability is currently being pursued through several different approaches, such as with Tiamat, which is a simulation tool within VERA that more tightly couples the MPACT neutron transport solver, the CTF thermal hydraulics solver, and the MOOSE-based Bison-CASL fuel performance code. However, the process in this papermore » focuses on running fuel performance calculations with Bison-CASL to predict PCI using the multicycle output data from coupled neutron transport/thermal hydraulics simulations. In recent work within CASL, Watts Bar Unit 1 has been simulated over 12 cycles using the VERA core simulator capability based on MPACT and CTF. Using the output from these simulations, Bison-CASL results can be obtained without rerunning all 12 cycles, while providing some insight into PCI indicators. Multi-cycle Bison-CASL results are presented and compared against results from the FRAPCON fuel performance code. There are several quantities of interest in considering PCI and subsequent fuel rod failures, such as the clad hoop stress and maximum centerline fuel temperature, particularly as a function of time. Bison-CASL performs single-rod simulations using representative power and temperature distributions, providing high-resolution results for these and a number of other quantities. This will assist in identifying fuels rods as potential failure locations for use in further analyses.« less

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

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

IJ van Rooyen; WR Lloyd; TL Trowbridge

2013-09-01

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designsmore » being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.« less

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

NASA Technical Reports Server (NTRS)

Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

1973-01-01

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Apparatuses for prestressing rod-type specimens in torsion for in-situ passive fracture toughness testing in an extremely high-pressure environment of hydrogen

DOEpatents

Wang, Jy-an [Oak Ridge, TN; Liu, Ken C [Oak Ridge, TN; Feng, Zhili [Knoxville, TN

2012-05-15

An in-situ specimen fixture particularly adapted for prestressing rod-type SNTT-type specimens comprising a tube and end cap wherein the specimen is secured at one end to the tube, and at the opposite end to the end cap. The end cap is rotatable relative to the tube, and may be fixedly secured for creating a torsional force prestressing the specimen enclosed within the tube.

High Resolution Neutron Radiography and Tomography of Hydrided Zircaloy-4 Cladding Materials

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

Smith, Tyler S; Bilheux, Hassina Z; Ray, Holly B

2015-01-01

Neutron radiography for hydrogen analysis was performed with several Zircaloy-4 cladding samples with controlled hydrogen concentrations up to 1100 ppm. Hydrogen charging was performed in a process tube that was heated to facilitate hydrogen absorption by the metal. A correlation between the hydrogen concentration in the hydrided tubes and the neutron intensity was established, by which hydrogen content can be determined precisely in a small area (55 m x 55 m). Radiography analysis was also performed to evaluate the heating rate and its correlation with the hydrogen distribution through hydrided materials. In addition to radiography analysis, tomography experiments were performedmore » on Zircaloy-4 tube samples to study the local hydrogen distribution. Through tomography analysis a 3D reconstruction of the tube was evaluated in which an uneven hydrogen distribution in the circumferential direction can be observed.« less

Preliminary posttest analysis of LOFT loss-of-coolant experiment L2-2

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

White, J.R.; Grush, W.H.; Keeler, C.D.

A preliminary posttest analysis of Loss-of-Coolant Experiment (LOCE) L2-2, which was conducted in the Loss-of-Fluid Test (LOFT) facility, was performed to gain an understanding of the cause of the disparity between predicted and measured fuel rod cladding temperature responses in the LOFT core. LOCE L2-2 is the first experiment in the LOFT Power Ascension Series L2 (first series of LOFT nuclear experiments), which was designed to investigate the response of the LOFT nuclear core to the blowdown, refill, and reflood transients during LOCEs conducted at gradually increasing power levels. LOCE L2-2 was conducted at 50% power (25 MW, 26.38 kW/m).more » Results show that a core-wide rewet occurred early in the transient (during blowdown starting at about 7 s after rupture) which was not calculated in the pretest prediction analysis. This early core-wide rewet resulted in the peak fuel rod cladding temperatures being lower (by a mean value of 166/sup 0/K for 24 thermocouples) than had been calculated. This preliminary posttest analysis was concerned solely with determining why the early core-wide rewet was not predicted by the RELAP4/MOD6 pretest analysis and be no means is it a complete posttest analysis of LOCE L2-2 results. However, during this analysis, several errors made in the prettest analysis were found, and their impact on the predicted results is assessed. Three factors were postulated to have caused the disparity between predicted and measured fuel rod cladding temperatures for LOCE L2-2: (a) the initial fuel rod stored energy, (b) the heat transfer surface, and (c) the hydraulics calculation. These factors were examined and are discussed in this report. It was determined that core hydraulics, as influenced by the calculation of broken loop cold leg break flow, was the major factor causing the disparity.« less

Uniaxial, Pure Bending, and Column Buckling Experiments on Superelastic NiTi Rods and Tubes

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

Watkins, Ryan T.; Reedlunn, Benjamin; Daly, Samantha

Many existing shape memory alloy (SMA) devices consist of slender beams and frames. To better understand SMA beam behavior, we experimentally examined the isothermal, room temperature response of superelastic NiTi rods and tubes, of similar outer diameters, subjected to four different modes of loading. Pure tension, pure compression, and pure bending experiments were first performed to establish and compare the baseline uniaxial and bending behaviors of rods and tubes. Column buckling experiments were then performed on rod and tube columns of several slenderness ratios to investigate their mechanical responses, phase transformation kinetics under combined uniaxial and bending deformation, and themore » interaction between material and structural instabilities. In all experiments, stereo digital image correlation measured local displacement fields in order to capture phenomena such as strain localization and propagating phase boundaries. Superelastic mechanical behavior and the nature of stress-induced phase transformation were found to be strongly affected by specimen geometry and the deformation mode. Under uniaxial tension, both the rod and tube had well-defined loading and unloading plateaus in their superelastic responses, during which stress-induced phase transformation propagated along the length of the specimen in the form of a high/low strain front. Due to the dependence of strain localization on kinematic compatibility, the high/low strain front morphologies differed between the rod and tube: for the rod, the high/low strain front consisted of a diffuse “neck”, while the high/low strain front in the tube consisted of distinct, criss-crossing “fingers.” During uniaxial compression, both cross-sectional forms exhibited higher transformation stresses and smaller transformation strains than uniaxial tension, highlighting the now well-known tension-compression asymmetry of SMAs. Additionally, phase transformation localization and propagation were absent under compressive loading. During pure bending, the moment-curvature response of both forms exhibited plateaus and strain localization during forward and reverse transformations. Rod specimens developed localized, high-curvature regions that propagated along the specimen axis and caused shear strain near the high/low curvature interface; whereas, the tube specimens exhibited finger/wedge-like high strain regions over the tensile side of the tube which caused nonlinear strain profiles through the thickness of the specimen that did not propagate. Here, it was therefore found that classical beam theory assumptions did not hold in the presence of phase transformation localization (although, the assumptions did hold on average for the tube). During column buckling, the structures were loaded into the post-buckling regime yet recovered nearly-straight forms upon unloading. Strain localization was observed only for high aspect ratio (slender) tubes, but the mechanical responses were similar to that of rods of the same slenderness ratio. Also, an interesting “unbuckling” phenomenon was discovered in certain low aspect ratio (stout) columns, where late post-buckling straightening was observed despite continuous monotonic loading. Thus, these behaviors are some of the challenging phenomena which must be captured when developing SMA constitutive models and executing structural simulations.« less

Uniaxial, Pure Bending, and Column Buckling Experiments on Superelastic NiTi Rods and Tubes

DOE PAGES

Watkins, Ryan T.; Reedlunn, Benjamin; Daly, Samantha; ...

2018-03-23

Many existing shape memory alloy (SMA) devices consist of slender beams and frames. To better understand SMA beam behavior, we experimentally examined the isothermal, room temperature response of superelastic NiTi rods and tubes, of similar outer diameters, subjected to four different modes of loading. Pure tension, pure compression, and pure bending experiments were first performed to establish and compare the baseline uniaxial and bending behaviors of rods and tubes. Column buckling experiments were then performed on rod and tube columns of several slenderness ratios to investigate their mechanical responses, phase transformation kinetics under combined uniaxial and bending deformation, and themore » interaction between material and structural instabilities. In all experiments, stereo digital image correlation measured local displacement fields in order to capture phenomena such as strain localization and propagating phase boundaries. Superelastic mechanical behavior and the nature of stress-induced phase transformation were found to be strongly affected by specimen geometry and the deformation mode. Under uniaxial tension, both the rod and tube had well-defined loading and unloading plateaus in their superelastic responses, during which stress-induced phase transformation propagated along the length of the specimen in the form of a high/low strain front. Due to the dependence of strain localization on kinematic compatibility, the high/low strain front morphologies differed between the rod and tube: for the rod, the high/low strain front consisted of a diffuse “neck”, while the high/low strain front in the tube consisted of distinct, criss-crossing “fingers.” During uniaxial compression, both cross-sectional forms exhibited higher transformation stresses and smaller transformation strains than uniaxial tension, highlighting the now well-known tension-compression asymmetry of SMAs. Additionally, phase transformation localization and propagation were absent under compressive loading. During pure bending, the moment-curvature response of both forms exhibited plateaus and strain localization during forward and reverse transformations. Rod specimens developed localized, high-curvature regions that propagated along the specimen axis and caused shear strain near the high/low curvature interface; whereas, the tube specimens exhibited finger/wedge-like high strain regions over the tensile side of the tube which caused nonlinear strain profiles through the thickness of the specimen that did not propagate. Here, it was therefore found that classical beam theory assumptions did not hold in the presence of phase transformation localization (although, the assumptions did hold on average for the tube). During column buckling, the structures were loaded into the post-buckling regime yet recovered nearly-straight forms upon unloading. Strain localization was observed only for high aspect ratio (slender) tubes, but the mechanical responses were similar to that of rods of the same slenderness ratio. Also, an interesting “unbuckling” phenomenon was discovered in certain low aspect ratio (stout) columns, where late post-buckling straightening was observed despite continuous monotonic loading. Thus, these behaviors are some of the challenging phenomena which must be captured when developing SMA constitutive models and executing structural simulations.« less

Method of forming magnetostrictive rods from rare earth-iron alloys

DOEpatents

McMasters, O.D.

1986-09-02

Rods of magnetostrictive alloys of iron with rare earth elements are formed by flowing a body of rare earth-iron alloy in a crucible enclosed in a chamber maintained under an inert gas atmosphere, forcing such molten rare-earth-iron alloy into a hollow mold tube of refractory material positioned with its lower end portion within the molten body by means of a pressure differential between the chamber and mold tube and maintaining a portion of the molten alloy in the crucible extending to a level above the lower end of the mold tube so that solid particles of higher melting impurities present in the alloy collect at the surface of the molten body and remain within the crucible as the rod is formed in the mold tube. 5 figs.

Method of forming magnetostrictive rods from rare earth-iron alloys

DOEpatents

McMasters, O. Dale

1986-09-02

Rods of magnetrostructive alloys of iron with rare earth elements are formed by flowing a body of rare earth-iron alloy in a crucible enclosed in a chamber maintained under an inert gas atmosphere, forcing such molten rare-earth-iron alloy into a hollow mold tube of refractory material positioned with its lower end portion within the molten body by means of a pressure differential between the chamber and mold tube and maintaining a portion of the molten alloy in the crucible extending to a level above the lower end of the mold tube so that solid particles of higher melting impurities present in the alloy collect at the surface of the molten body and remain within the crucible as the rod is formed in the mold tube.

Self-monitoring high voltage transmission line suspension insulator

DOEpatents

Stemler, Gary E.; Scott, Donald N.

1981-01-01

A high voltage transmission line suspension insulator (18 or 22) which monitors its own dielectric integrity. A dielectric rod (10) has one larger diameter end fitting attachable to a transmission line and another larger diameter end fitting attachable to a support tower. The rod is enclosed in a dielectric tube (14) which is hermetically sealed to the rod's end fittings such that a liquidtight space (20) is formed between the rod and the tube. A pressurized dielectric liquid is placed within that space. A discoloring dye placed within this space is used to detect the loss of the pressurized liquid.

Tapered fiber Mach-Zehnder interferometers for vibration and elasticity sensing applications.

PubMed

Chen, Nan-Kuang; Hsieh, Yu-Hsin; Lee, Yi-Kun

2013-05-06

We demonstrate the optical measurements of heart-beat pulse rate and also elasticity of a polymeric tube, using a tapered fiber Mach-Zehnder interferometer. This device has two abrupt tapers in the Er/Yb codoped fiber and thus fractional amount of core mode is converted into cladding modes at the first abrupt taper. The core and cladding modes propagate through different optical paths and meet again at the second abrupt taper to produce interferences. The mechanical vibration signals generated by the blood vessels and by an inflated polymeric tube can perturb the optical paths of resonant modes to move around the resonant wavelengths. Thus, the cw laser signal is modulated to become pulses to reflect the heart-beat pulse rate and the elasticity of a polymeric tube, respectively.

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

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

Gamble, K. A.; Hales, J. D.; Zhang, Y.

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. Thismore » allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to utilize lower length scale approaches (e.g., density functional theory, cluster dynamics, rate theory, phase field, and Visco-Plastic- Self-Consistent (VPSC)) to develop more physically informed models at the engineering scale for use in the BISON [9] fuel performance code.« less

Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

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

Yang, Yong; Phillpot, Simon

Fuel cladding chemical interactions (FCCI) have been acknowledged as a critical issue in a metallic fuel/steel cladding system due to the formation of low melting intermetallic eutectic compounds between the fuel and cladding steel, resulting in reduction in cladding wall thickness as well as a formation of eutectic compounds that can initiate melting in the fuel at lower temperature. In order to mitigate FCCI, diffusion barrier coatings on the cladding inner surface have been considered. In order to generate the required coating techniques, pack cementation, electroplating, and electrophoretic deposition have been investigated. However, these methods require a high processing temperaturemore » of above 700 oC, resulting in decarburization and decomposition of the martensites in a ferritic/martensitic (F/M) cladding steel. Alternatively, organometallic chemical vapor deposition (OMCVD) can be a promising process due to its low processing temperature of below 600 oC. The aim of the project is to conduct applied and fundamental research towards the development of diffusion barrier coatings on the inner surface of F/M fuel cladding tubes. Advanced cladding steels such as T91, HT9 and NF616 have been developed and extensively studied as advanced cladding materials due to their excellent irradiation and corrosion resistance. However, the FCCI accelerated by the elevated temperature and high neutron exposure anticipated in fast reactors, can have severe detrimental effects on the cladding steels through the diffusion of Fe into fuel and lanthanides towards into the claddings. To test the functionality of developed coating layer, the diffusion couple experiments were focused on using T91 as cladding and Ce as a surrogate lanthanum fission product. By using the customized OMCVD coating equipment, thin and compact layers with a few micron between 1.5 µm and 8 µm thick and average grain size of 200 nm and 5 µm were successfully obtained at the specimen coated between 300oC and 500 oC, respectively. The coating layer contains both carbon and vanadium elements as quantified by WED, and the phases mainly consist of a mixture of V2C and VC, which was confirmed using X-ray diffraction patterns. In addition, the ratio between V and C varies with processing temperature, and it was observed that a higher temperature promotes the carbon adsorption and increases thickness of the coating. With optimized deposition conditions, we can apply the coating technique toward the actual T91 cladding materials, and provide the possibilities for the real application in sodium-cooled fast reactors (SFRs). Diffusion couple experiments were performed at both 550 oC and 690 oC, which corresponds to normal and aggressive operating temperatures, respectively. The results show that vanadium carbide coating with wider thickness (8 µm) and lower carbon concentration (27 at.%) reduced the width of the inter diffusion region, indicating that vanadium carbide coating can mitigate FCCI effectively. In specific, inter-diffusion between Fe and Ce was prohibited over most area, but Ce diffusion occurred toward the coating and the Fe substrate through thinner coating layer, which needs further optimization in terms of uniform coating thickness. Overall, it is concluded that this coating process can be successfully applied onto the inner surface of HT9 cladding tubes and the FCCI can be effectively mitigated if not totally eliminated.« less

Bubble snap-off and capillary-back pressure during counter-current spontaneous imbibition into model pores.

PubMed

Unsal, Evren; Mason, Geoffrey; Morrow, Norman R; Ruth, Douglas W

2009-04-09

A previous paper (Unsal, E.; Mason, G.; Ruth, D. W.; Morrow, N. R. J. Colloid Interface Sci. 2007, 315, 200-209) reported experiments involving counter-current spontaneous imbibition into a model pore system consisting of a rod in an angled slot covered by a glass plate. Such an arrangement gives two tubes with different cross-sections (both size and shape) with an interconnection through the gap between the rod and the plate. In the previous experiments, the wetting phase advanced in the small tube and nonwetting phase retreated in the large tube. No bubbles were formed. In this paper, we study experimentally and theoretically the formation of bubbles at the open end of the large tube and their subsequent snap-off. Such bubbles reduce the capillary back pressure produced by the larger tube and can thus have an effect on the local rate of imbibition. In the model pore system, the rod was either in contact with the glass, forming two independent tubes, or the rod was spaced from the glass to allow cross-flow between the tubes. For small gaps, there were three distinct menisci. The one with the highest curvature was between the rod and the plate. The next most highly curved was in the smaller tube, and the least highly curved meniscus was in the large tube and this was the tube from which the bubbles developed. The pressure in the dead end of the system was recorded during imbibition. Once the bubble starts to form outside of the tube, the pressure drops rapidly and then steadies. After the bubble snaps off, the pressure rises to almost the initial value and stays essentially constant until the next bubble starts to form. After snap-off, the meniscus in the large tube appears to invade the large tube for some distance. The snap-off is the result of capillary instability; it takes place significantly inside the large tube with flow of wetting phase moving in the angular corners. As imbibition into the small tube progresses, the rate of imbibition decreases and the time taken for each bubble to form increases, slightly increasing the pressure at which snap-off occurs. The snap-off curvature is only about two-thirds of the curvature of a theoretical cylindrical meniscus within the large tube and about 40% of the curvature of the actual meniscus spanning the large tube.

The Influence of the In-Situ Clad Staining on the Corrosion of Zircaloy in PWR Water Environment

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

Kammenzind, B.F., Eklund, K.L. and Bajaj, R.

Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the diametrical direction under the external pressure of the water environment, but potentially grow later in life in the diametrical direction once the expanding fuel pelletmore » contacts the cladding inner wall (Reference (b)). Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction (Reference (c)). Once above the solubility limit in Zircaloy, the hydride precipitates as zirconium hydride (References (c) through (j)). Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal (Reference (k)). Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface (References (l) through (n)), in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer. References (o) through (q) investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference (o), reported that the applied tensile stress increased corrosion rates above a critical stress level in 400 C and 475 C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed stress effect. Kim et al. (Reference (p)) and Kim and Kim (Reference (q)) more recently investigated the influence that an applied hoop stress has on the corrosion resistance of Zircaloy tubes in a 400 C steam and in a 350 C concentrated lithia water environment. Both of these studies found the applied tensile hoop stress to have no effect on cladding corrosion rates in the 400 C steam environment but to have accelerated corrosion in the lithiated water environment. In both cases, the corrosion acceleration in the lithiated water environment was attributed to the accumulation of the increased hydrogen picked up in the lithiated environment into the tensile regions of the test specimen. Dense hydride rims have been shown, independent of clad strain, to accelerate the corrosion of Zirconium alloys (References (r) and (s)), suggesting that the primary effect of applied stresses on the corrosion of Zircaloy in the above studies is through the accumulation of hydrogen at the oxide-to-metal interface and not through a direct mechanical breakdown of the passivating boundary layer. To further investigate the potential role of in-situ clad straining (or stress) on Zircaloy corrosion rates, two experimental studies were performed. First, several samples that were irradiated with and without an applied stress were destructively examined for the extent of corrosion occurring in strained and nonstrained regions of the test samples. The extent of corrosion was determined, posttest, by metallographic examination. Second, the corrosion process was monitored in-situ using electrochemical impedance spectroscopy on samples exposed out-of-reactor with and without an applied stress. Post test, these autoclave samples were also metallographically examined.« less

Perform Tests and Document Results and Analysis of Oxide Layer Effects and Comparisons

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

Collins, E. D.; DelCul, G. D.; Spencer, B. B.

2014-08-30

During the initial feasibility test using actual used nuclear fuel (UNF) cladding in FY 2012, an incubation period of 30–45 minutes was observed in the initial dry chlorination. The cladding hull used in the test had been previously oxidized in a dry air oxidation pretreatment prior to removal of the fuel. The cause of this incubation period was attributed to the resistance to chlorination of an oxide layer imparted by the dry oxidation pretreatment on the cladding. Subsequently in 2013, researchers at the Korea Atomic Energy Institute (KAERI) reported on their chlorination study [R1] on ~9-gram samples of unirradiated ZirloTMmore » cladding tubes that had been previously oxidized in air at 500oC for various time periods to impart oxide layers of varying thickness. In early 2014, discussions with Indefinite Delivery, Indefinite Quantity (IDIQ) contracted technical consultants from Westinghouse described their previous development (and patents) [R2] on methods of chemical washing to remove some or all of the hydrous oxide layer imparted on UNF cladding during irradiation in light water reactors (LWRs) . Thus, the Oak Ridge National Laboratory (ORNL) study, described herein, was planned to extend the KAERI study on the effects of anhydrous oxide layers, but on larger ~100-gram samples of unirradiated zirconium alloy cladding tubes, and to investigate the effects of various methods of chemical pretreatment prior to chlorination with 100% chlorine on the average reaction rates and Cl2 usage efficiencies.« less

Comminuting irradiated ferritic steel

DOEpatents

Bauer, Roger E.; Straalsund, Jerry L.; Chin, Bryan A.

1985-01-01

Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

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.

Development of modified MDA (M-MDA), PWR fuel cladding tube for high duty operation in future

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

Watanabe, Seiichi; Kido, Toshiya; Arakawa, Yasushi

2007-07-01

A new cladding material of M-MDA has been developed in order to prepare for a strong growing demand for advanced fuel which can maintain its integrity even under high duties due to more efficient operation such as higher burnup, higher LHR, and longer operation cycle which will contribute the suppression of environmental burdens like CO{sub 2} emission. The main aim of M-MDA is to have excellent corrosion resistance while the other properties are inherited from MDA, which has been adopted to the step 2 fuel, instead of Zry-4, of Japanese PWR plant whose upper limit of assembly discharged burnup ismore » 55 MWd/kgU. And we could confirm that the main aim of M-MDA was achieved by means of out-of-pile tests. In order to confirm improvement of corrosion resistance of M-MDA in the actual operation, irradiation test of M-MDA in the commercial reactor of Vandellos II is ongoing. The latest results of on-site examination after every end of cycle showed that oxide thickness of M-MDA-SR was much smaller than that of MDA at rod discharged burnup of approximately 60 MWd/kgU. The final irradiation cycle was completed on April 2007 and then we will obtain corrosion data of M-MDA over 70 MWd/kgU. M-MDA is a candidate alloy for advanced fuel under higher duty usage. (authors)« less

Controlled fragmentation of multimaterial fibres and films via polymer cold-drawing.

PubMed

Shabahang, Soroush; Tao, Guangming; Kaufman, Joshua J; Qiao, Yangyang; Wei, Lei; Bouchenot, Thomas; Gordon, Ali P; Fink, Yoel; Bai, Yuanli; Hoy, Robert S; Abouraddy, Ayman F

2016-06-23

Polymer cold-drawing is a process in which tensile stress reduces the diameter of a drawn fibre (or thickness of a drawn film) and orients the polymeric chains. Cold-drawing has long been used in industrial applications, including the production of flexible fibres with high tensile strength such as polyester and nylon. However, cold-drawing of a composite structure has been less studied. Here we show that in a multimaterial fibre composed of a brittle core embedded in a ductile polymer cladding, cold-drawing results in a surprising phenomenon: controllable and sequential fragmentation of the core to produce uniformly sized rods along metres of fibre, rather than the expected random or chaotic fragmentation. These embedded structures arise from mechanical-geometric instabilities associated with 'neck' propagation. Embedded, structured multimaterial threads with complex transverse geometry are thus fragmented into a periodic train of rods held stationary in the polymer cladding. These rods can then be easily extracted via selective dissolution of the cladding, or can self-heal by thermal restoration to re-form the brittle thread. Our method is also applicable to composites with flat rather than cylindrical geometries, in which case cold-drawing leads to the break-up of an embedded or coated brittle film into narrow parallel strips that are aligned normally to the drawing axis. A range of materials was explored to establish the universality of this effect, including silicon, germanium, gold, glasses, silk, polystyrene, biodegradable polymers and ice. We observe, and verify through nonlinear finite-element simulations, a linear relationship between the smallest transverse scale and the longitudinal break-up period. These results may lead to the development of dynamical and thermoreversible camouflaging via a nanoscale Venetian-blind effect, and the fabrication of large-area structured surfaces that facilitate high-sensitivity bio-detection.

Controlled fragmentation of multimaterial fibres and films via polymer cold-drawing

NASA Astrophysics Data System (ADS)

Shabahang, Soroush; Tao, Guangming; Kaufman, Joshua J.; Qiao, Yangyang; Wei, Lei; Bouchenot, Thomas; Gordon, Ali P.; Fink, Yoel; Bai, Yuanli; Hoy, Robert S.; Abouraddy, Ayman F.

2016-06-01

Polymer cold-drawing is a process in which tensile stress reduces the diameter of a drawn fibre (or thickness of a drawn film) and orients the polymeric chains. Cold-drawing has long been used in industrial applications, including the production of flexible fibres with high tensile strength such as polyester and nylon. However, cold-drawing of a composite structure has been less studied. Here we show that in a multimaterial fibre composed of a brittle core embedded in a ductile polymer cladding, cold-drawing results in a surprising phenomenon: controllable and sequential fragmentation of the core to produce uniformly sized rods along metres of fibre, rather than the expected random or chaotic fragmentation. These embedded structures arise from mechanical-geometric instabilities associated with ‘neck’ propagation. Embedded, structured multimaterial threads with complex transverse geometry are thus fragmented into a periodic train of rods held stationary in the polymer cladding. These rods can then be easily extracted via selective dissolution of the cladding, or can self-heal by thermal restoration to re-form the brittle thread. Our method is also applicable to composites with flat rather than cylindrical geometries, in which case cold-drawing leads to the break-up of an embedded or coated brittle film into narrow parallel strips that are aligned normally to the drawing axis. A range of materials was explored to establish the universality of this effect, including silicon, germanium, gold, glasses, silk, polystyrene, biodegradable polymers and ice. We observe, and verify through nonlinear finite-element simulations, a linear relationship between the smallest transverse scale and the longitudinal break-up period. These results may lead to the development of dynamical and thermoreversible camouflaging via a nanoscale Venetian-blind effect, and the fabrication of large-area structured surfaces that facilitate high-sensitivity bio-detection.

Effects of hydrogen on thermal creep behaviour of Zircaloy fuel cladding

NASA Astrophysics Data System (ADS)

Suman, Siddharth; Khan, Mohd Kaleem; Pathak, Manabendra; Singh, R. N.

2018-01-01

Zirconium alloys are extensively used for nuclear fuel cladding. Creep is one of the most likely degradation mechanisms for fuel cladding during reactor operating and repository conditions. Fuel cladding tubes undergo waterside corrosion during service and hydrogen is produced as a result of it-a fraction of which is picked up by cladding. Hydrogen remains in solid solution up to terminal solid solubility and it precipitates as brittle hydride phase in the zirconium metal matrix beyond this limiting concentration. Hydrogen, either in solid solution or as precipitated hydride, alters the creep behaviour of zirconium alloys. The present article critically reviews the influence of hydrogen on thermal creep behaviour of zirconium alloys, develops the systematic understanding of this multifaceted phenomenon, and delineates the thrust areas which require further investigations.

Process development and fabrication of space station type aluminum-clad graphite epoxy struts

NASA Technical Reports Server (NTRS)

Ring, L. R.

1990-01-01

The manufacture of aluminum-clad graphite epoxy struts, designed for application to the Space Station truss structure, is described. The strut requirements are identified, and the strut material selection rationale is discussed. The manufacturing procedure is described, and shop documents describing the details are included. Dry graphite fiber, Pitch-75, is pulled between two concentric aluminum tubes. Epoxy resin is then injected and cured. After reduction of the aluminum wall thickness by chemical milling the end fittings are bonded on the tubes. A discussion of the characteristics of the manufactured struts, i.e., geometry, weight, and any anomalies of the individual struts is included.

Tensile properties and flow behavior analysis of modified 9Cr-1Mo steel clad tube material

NASA Astrophysics Data System (ADS)

Singh, Kanwarjeet; Latha, S.; Nandagopal, M.; Mathew, M. D.; Laha, K.; Jayakumar, T.

2014-11-01

The tensile properties and flow behavior of modified 9Cr-1Mo steel clad tube have been investigated in the framework of various constitutive equations for a wide range of temperatures (300-923 K) and strain rates (3 × 10-3 s-1, 3 × 10-4 s-1 and 3 × 10-5 s-1). The tensile flow behavior of modified 9Cr-1Mo steel clad tube was most accurately described by Voce equation. The variation of instantaneous work hardening rate (θ = dσ/dε) and σθ with stress (σ) indicated two stage behavior characterized by rapid decrease at low stresses (transient stage) followed by a gradual decrease in high stresses (Stage III). The variation of work hardening parameters and work hardening rate in terms of θ vs. σ and σθ vs. σ with temperature exhibited three distinct regimes. Rapid decrease in flow stress and work hardening parameters and rapid shift of θ vs. σ and σθ vs. σ towards low stresses with increase in temperature indicated dynamic recovery at high temperatures. Tensile properties of the material have been best predicted from Voce equation.

CONTROL ROD

DOEpatents

Walker, D.E.; Matras, S.

1963-04-30

This patent shows a method of making a fuel or control rod for a nuclear reactor. Fuel or control material is placed within a tube and plugs of porous metal wool are inserted at both ends. The metal wool is then compacted and the tube compressed around it as by swaging, thereby making the plugs liquid- impervious but gas-pervious. (AEC)

77 FR 36137 - Airworthiness Directives; AGUSTA S.p.A. Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-18

... the upper end of collective control rod C2 to torque tube C3 is properly installed. This AD is... prevent separation of the collective control rod from the torque tube, loss of control of the collective... helicopters because the production quality control procedures did not require recording the applied torque on...

Texture control of zircaloy tubing during tube reduction

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

Nagai, N.; Kakuma, T.; Fujita, K.

1982-01-01

Seven batches of Zircaloy-2 nuclear fuel cladding tubes with different textures were processed from tube shells of the same size, by different reduction routes, using pilger and 3-roll mills. Based on the texture data of these tubes, the texture control of Zircaloy tubing, the texture gradient across the wall, and the texture change during annealing were studied. The deformation texture of Zicaloy-2 tubing was dependent on the tool's curvature and was independent of the dimensions of the mother tubes. The different slopes of texture gradients were observed between the tubing of higher strain ration and that of lower strain ratio.

Technique development for modulus, microcracking, hermeticity, and coating evaluation capability characterization of SiC/SiC tubes

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

Hu, Xunxiang; Ang, Caen K.; Singh, Gyanender P.

Driven by the need to enlarge the safety margins of nuclear fission reactors in accident scenarios, research and development of accident-tolerant fuel has become an important topic in the nuclear engineering and materials community. A continuous-fiber SiC/SiC composite is under consideration as a replacement for traditional zirconium alloy cladding owing to its high-temperature stability, chemical inertness, and exceptional irradiation resistance. An important task is the development of characterization techniques for SiC/SiC cladding, since traditional work using rectangular bars or disks cannot directly provide useful information on the properties of SiC/SiC composite tubes for fuel cladding applications. At Oak Ridge Nationalmore » Laboratory, experimental capabilities are under development to characterize the modulus, microcracking, and hermeticity of as-fabricated, as-irradiated SiC/SiC composite tubes. Resonant ultrasound spectroscopy has been validated as a promising technique to evaluate the elastic properties of SiC/SiC composite tubes and microcracking within the material. A similar technique, impulse excitation, is efficient in determining the basic mechanical properties of SiC bars prepared by chemical vapor deposition; it also has potential for application in studying the mechanical properties of SiC/SiC composite tubes. Complete evaluation of the quality of the developed coatings, a major mitigation strategy against gas permeation and hydrothermal corrosion, requires the deployment of various experimental techniques, such as scratch indentation, tensile pulling-off tests, and scanning electron microscopy. In addition, a comprehensive permeation test station is being established to assess the hermeticity of SiC/SiC composite tubes and to determine the H/D/He permeability of SiC/SiC composites. This report summarizes the current status of the development of these experimental capabilities.« less

Fatigue of Ti-3A1-2.5V Alloy Tube and Rod

DTIC Science & Technology

2007-02-28

during cyclic straining of a 18% nickel maraging steel and attributed it to the presence of a crack. Skelton (reference 6) also attributed a hysteresis...growth, and fracture lives was also defined. The LCF behavior of the alloy tube and rod was investigated, examining the shape change of load...is weldable as the commercially pure grades and has excellent resistance to torsion and corrosion . Therefore, it is used principally as tubing in

Structure-dependent performance of TiO 2/C as anode material for Na-ion batteries

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

He, Hanna; Gan, Qingmeng; Wang, Haiyan

The performance of energy storage materials is highly dependent on their nanostructures. Herein, hierarchical rod-in-tube TiO 2 with a uniform carbon coating is synthesized as the anode material for sodium-ion batteries by a facile solvothermal method. This unique structure consists of a tunable nanorod core, interstitial hollow spaces, and a functional nanotube shell assembled from two-dimensional nanosheets. By adjusting the types of solvents used and reaction time, the morphologies of TiO 2/C composites can be tuned to nanoparticles, microrods, rod-in-tube structures, or microtubes. Among these materials, rod-in-tube TiO 2 with a uniform carbon coating shows the highest electronic conductivity, specificmore » surface area (336.4 m(2) g(-1)), and porosity, and these factors lead to the best sodium storage capability. Benefiting from the unique structural features and improved electronic/ionic conductivity, the as-obtained rod-in-tube TiO2/C in coin cell tests exhibits a high discharge capacity of 277.5 and 153.9 mAh g(-1) at 50 and 5000 mA g(-1), respectively, and almost 100% capacity retention over 14,000 cycles at 5000 mA g(-1). In operando high-energy X-ray diffraction further confirms the stable crystal structure of the rod-in-tube TiO 2/C during Na+ insertion/extraction. This work highlights that nanostructure design is an effective strategy to achieve advanced energy storage materials.« less

Seeding of capsule instability growth by fill tubes and support rods for inertial confinement fusion implosions

NASA Astrophysics Data System (ADS)

Macphee, Andrew; Casey, Daniel; Clark, Daniel; Field, John; Haan, Steven; Hammel, Bruce; Kroll, Jeremy; Landen, Otto; Martinez, David; Milovich, Jose; Nikroo, Abbas; Rice, Neal; Robey, Harry; Smalyuk, Vladimir; Stadermann, Michael; Weber, Christopher; Lawrence Livermore National Laboratory Collaboration; Atomics Collaboration, General

2016-10-01

Features associated with the target support tent and deuterium-tritium fuel fill tube and support rods can seed hydrodynamic instabilities leading to degraded performance for inertial confinement fusion (ICF) experiments at the National Ignition Facility. We performed in-flight radiography of ICF capsules in the vicinity of the capsule support tent and fill tube surrogates to investigate instability growth associated with these features. For both plastic and high density carbon ablators, the shadow of the 10 μm diameter glass fill-tube cast by the x-ray spots on the hohlraum wall were observed to imprint radial instabilities around the fill tube/capsule interface. Similarly, instability growth was observed for the shadow cast by 12 μm diameter silicon carbide capsule support rods mounted orthogonal to the fill tube as a tent alternative for a plastic ablator. The orientation of the shadows is consistent with raytracing. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Real-time monitoring of barrel thickness and barrel/screw separation using ultrasound

NASA Astrophysics Data System (ADS)

Jen, Cheng-Kuei; Zun, Zhigang; Kobayashi, Makiko

2005-03-01

Ultrasonic sensors together with a fast data acquisition system have been used to monitor the barrel thickness and barrel/screw separation during low-density polyethylene as well as high-density polyethylene extrusion in 30 mm and 50 mm twin-screw extruders. The sensors include sol-gel sprayed high temperature (HT) piezoelectric thick ceramic film ultrasonic transducers (UTs), stand-alone HTUTs and air-cooled buffer rod type sensors consisting of a room temperature UT and a non-clad or clad buffer rod to which the room temperature UT is attached. The installation and use of these sensors are non-intrusive to the extruder and non-destructive to the polymers being processed. This study has demonstrated the capability of appropriately designed ultrasonic sensors in monitoring the barrel and screw integrity at the melting, mixing and pumping zones of the extruder via barrel or flange. The merits and limitations of these sensors are discussed. The measurement speed and analysis of the sensitivity for quantitative wear measurements are also presented.

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

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

R. L. Williamson

A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete andmore » smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.« less

Characteristics of copper-clad aluminum rods prepared by horizontal continuous casting

NASA Astrophysics Data System (ADS)

Zhang, Yubo; Fu, Ying; Jie, Jinchuan; Wu, Li; Svynarenko, Kateryna; Guo, Qingtao; Li, Tingju; Wang, Tongmin

2017-11-01

An innovative horizontal continuous casting method was developed and successfully used to prepare copper-clad aluminum (CCA) rods with a diameter of 85 mm and a sheath thickness of 16 mm. The solidification structure and element distribution near the interface of the CCA ingots were investigated by means of a scanning electron microscope, an energy dispersive spectrometer, and an electron probe X-ray microanalyzer. The results showed that the proposed process can lead to a good metallurgical bond between Cu and Al. The interface between Cu and Al was a multilayered structure with a thickness of 200 μm, consisting of Cu9Al4, CuAl2, α-Al/CuAl2 eutectic, and α-Al + α-Al/CuAl2 eutectic layers from the Cu side to the Al side. The mean tensile-shear strength of the CCA sample was 45 MPa, which fulfills the requirements for the further extrusion process. The bonding and diffusion mechanisms are also discussed in this paper.

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

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

Saltzstein, Sylvia J.; Sorenson, Ken B.; Hanson, Brady

The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened andmore » the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.« less

Slip-spring model of entangled rod-coil block copolymers

NASA Astrophysics Data System (ADS)

Wang, Muzhou; Likhtman, Alexei E.; Olsen, Bradley D.

2015-03-01

Understanding the dynamics of rod-coil block copolymers is important for optimal design of functional nanostructured materials for organic electronics and biomaterials. Recently, we proposed a reptation theory of entangled rod-coil block copolymers, predicting the relaxation mechanisms of activated reptation and arm retraction that slow rod-coil dynamics relative to coil and rod homopolymers, respectively. In this work, we introduce a coarse-grained slip-spring model of rod-coil block copolymers to further explore these mechanisms. First, parameters of the coarse-grained model are tuned to match previous molecular dynamics simulation results for coils, rods, and block copolymers. For activated reptation, rod-coil copolymers are shown to disfavor configurations where the rod occupies curved portions of the entanglement tube of randomly varying curvature created by the coil ends. The effect of these barriers on diffusion is quantitatively captured by considering one-dimensional motion along an entanglement tube with a rough free energy potential. Finally, we analyze the crossover between the two mechanisms. The resulting dynamics from both mechanisms acting in combination is faster than from each one individually.

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

Copper vapor laser acoustic thermometry system

DOEpatents

Galkowski, Joseph J.

1987-01-01

A copper vapor laser (CVL) acoustic thermometry system is disclosed. The invention couples an acoustic pulse a predetermined distance into a laser tube by means of a transducer and an alumina rod such that an echo pulse is returned along the alumina rod to the point of entry. The time differential between the point of entry of the acoustic pulse into the laser tube and the exit of the echo pulse is related to the temperature at the predetermined distance within the laser tube. This information is processed and can provide an accurate indication of the average temperature within the laser tube.

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

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

Pastore, G.; Novascone, S. R.; Williamson, R. L.

2015-09-01

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rodmore » undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.« less

Methodology for Mechanical Property Testing of Fuel Cladding Using a Expanded Plug Wedge Test

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

Jiang, Hao; Wang, Jy-An John

2014-01-01

An expanded plug method was developed earlier for determining the tensile properties of irradiated fuel cladding. This method tests fuel rod cladding ductility by utilizing an expandable plug to radially stretch a small ring of irradiated cladding material. The circumferential or hoop strain is determined from the measured diametrical expansion of the ring. A developed procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves, from which material properties of the cladding can be extracted. However, several deficiencies existed in this expanded-plug test that can impact the accuracy of test results, suchmore » as that the large axial compressive stress resulted from the expansion plug test can potentially induce the shear failure mode of the tested specimen. Moreover, highly nonuniform stress and strain distribution in the deformed clad gage section and significant compressive stresses, induced by bending deformation due to clad bulging effect, will further result in highly nonconservative estimates of the mechanical properties for both strength and ductility of the tested clad. To overcome the aforementioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. By optimizing the specific geometry designs, selecting the appropriate material for the expansion plug, and adding new components into the testing system, a modified expansion plug testing protocol has been developed. A general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor, -factor, was used to convert the ring load Fring into hoop stress , and is written as _ = F_ring/tl , where t is the clad thickness and l is the clad length. The generated stress-strain curve agrees well with the associated tensile test data in both elastic and plastic deformation regions.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-25

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

Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

DOEpatents

Brehm, Jr., William F.; Colburn, Richard P.

1982-01-01

An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

NEUTRONIC REACTORS

DOEpatents

Anderson, H.L.

1958-10-01

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Overview of Fuel Rod Simulator Usage at ORNL

NASA Astrophysics Data System (ADS)

Ott, Larry J.; McCulloch, Reg

2004-02-01

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

Technical activities report: Heat, water, and mechanical studies

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

Alexander, W.K.

1951-10-04

Topics in the heat studies section include: front and rear face reflector shields at the C-pile; process tube channel thermocouples; water temperature limits for horizontal rods; slug temperature and thermal conductivity calculations; maximum slug-end cap temperature; boiling consideration studies; scram time limit for Panellit alarm; heat transfer test; slug stresses; thermal insulation of bottom tube row at C-pile; flow tests; present pile enrichment; electric analog; and measurement of thermal contact resistance. Topics in the water studies section include: 100-D flow laboratory; process water studies; fundamental studies on film formation; coatings on tip-offs; can difference tests; slug jacket abrasion at highmore » flow rates; corrosion studies; front tube dummy slugs; metallographic examination of tubes from H-pile; fifty-tube mock-up; induction heating facility; operational procedures and standards; vertical safety rod dropping time tests; recirculation; and power recovery. Mechanical development studies include: effect of Sphincter seal and lubricant VSR drop time; slug damage; slug bubble tester; P-13 removal; chemical slug stripper; effect of process tube rib spacing and width; ink facility installation; charging and discharging machines; process tube creep; flapper nozzle assembly test; test of single gun barrel assembly; pigtail fixture test; horizontal rod gland seal test; function test of C-pile; and intermediate test of Ball 3-X and VSR systems.« less

Silicone infusion tubing instead of Hunter rods for two-stage zone 2 flexor tendon reconstruction in a resource-limited surgical environment.

PubMed

Kibadi, K; Moutet, F

2017-10-01

The authors describe their experience using silicone infusion tubing in place of Hunter rods for two-stage zone 2 flexor tendon reconstruction in a resource-limited surgical environment. This case report features a 47-year-old, right-handed man who had no active PIP and DIP joint flexion in four fingers of the right hand 5 months after an injury. During the first repair stage, the A2 and A4 pulleys were reconstructed using an extensor retinaculum graft. An infusion tube was inserted instead of Hunter rods. During the second stage, formation of a digital neo-canal around the infusion tubing was observed. The infusion tubing was removed and replaced with a palmaris longus tendon graft according to the conventional technique. Physiotherapy and rehabilitation followed surgery. At 6 months, very significant progress had been made with complete recovery of PIP and DIP flexion in the four fingers. Copyright © 2017 SFCM. Published by Elsevier Masson SAS. All rights reserved.

Development of new ferritic steels as cladding material for metallic fuel fast breeder reactor

NASA Astrophysics Data System (ADS)

Tokiwai, Moriyasu; Horie, Masaaki; Kako, Kenji; Fujiwara, Masayuki

1993-09-01

The excellent thermal, chemical and neutronic properties of metallic fuel (U-Pu-Zr alloy) will lead to drastic improvements in fast reactor safety and the related fuel cycle economy. Some new high molybdenum 12Cr ferritic stainless steel candidate cladding alloys have been designed to achieve the mechanical properties required for high performance metallic fuel elements. These candidate claddings were irradiated by ion bombardment and tested to determine their strength and creep rupture properties. A 12Cr-8Mo and a 12Cr-8Mo-0.1Y 2O 3 steel were fabricated into cladding via a powder metallurgy process and by a mechanical alloying process, respectively. These claddings had two and three times the creep rupture strength (pressurized at 650°C for 10000 h) of a conventional 12Cr ferritic steel (HT-9). These two steels also showed no void formation up to 350 dpa by Ni 3+ irradiation. A zircaloy-2 lined steel cladding tube has also been fabricated for the purpose of reducing fuel-cladding interdiffusion and chemical interaction.

Electrothermal fracturing of tensile specimens

NASA Technical Reports Server (NTRS)

Blinn, H. O.; Hanks, J. G.; Perkins, H. P.

1970-01-01

Pulling device consisting of structural tube, connecting rod, spring-loaded nuts, loading rod, heating element, and three bulkheads fractures tensile specimens. Alternate heating and cooling increases tensile loading by increments until fracturing occurs. Load cell or strain gage, applied to pulling rod, determines forces applied.

Displacer Diameter Effect in Displacer Pulse Tube Refrigerator

NASA Astrophysics Data System (ADS)

Zhu, Shaowei

2017-12-01

Gas driving displacer pulse tube refrigerators are one of the work recovery type of pulse tube refrigerators whose theoretical efficiency is the same as Stirling refrigerators'. Its cooling power is from the displacement of the displacer. Displace diameter, rod diameter and pressure drop of the regenerator influence the displacement, which are investigated by numerical simulation. It is shown that the displacement ratio of the displacer over the piston is almost not affected by the displacer diameter at the same rod diameter ratio, or displacer with different diameters almost has the same performance.

Nuclear reactor remote disconnect control rod coupling indicator

DOEpatents

Vuckovich, Michael

1977-01-01

A coupling indicator for use with nuclear reactor control rod assemblies which have remotely disengageable couplings between the control rod and the control rod drive shaft. The coupling indicator indicates whether the control rod and the control rod drive shaft are engaged or disengaged. A resistive network, utilizing magnetic reed switches, senses the position of the control rod drive mechanism lead screw and the control rod position indicating tube, and the relative position of these two elements with respect to each other is compared to determine whether the coupling is engaged or disengaged.

RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg

NASA Astrophysics Data System (ADS)

Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo

Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

Solid-state-laser-rod holder

DOEpatents

Gettemy, D.J.; Barnes, N.P.; Griggs, J.E.

1981-08-11

The disclosure relates to a solid state laser rod holder comprising Invar, copper tubing, and epoxy joints. Materials and coefficients of expansion of the components of the holder combine with the rod to produce a joint which will give before the rod itself will. The rod may be lased at about 70 to 80/sup 0/K and returned from such a temperature to room temperature repeatedly without its or the holder's destruction.

Implementation of DoD ManTech Projects Receiving FY03-05 Funds

DTIC Science & Technology

2008-12-01

ManTech project is providing an integrated tool to grind, laser clad , and finish repair work of submarine VLS tubes. The system is scheduled for...hours from about 400 - 500 hours to 40 - 50 hours resulting in improved readiness. The in- situ laser clad process eliminates the use of hazardous...Alternate Remote Shipboard Lighting for Reduced Costs .......................................................... 14 2.2.10 ManTech for Military Lasers

Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

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

ORNL Interim Progress Report on Hydride Reorientation CIRFT Tests

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

Wang, Jy-An John; Yan, Yong; Wang, Hong

A systematic study of H. B. Robinson (HBR) high burnup spent nuclear fuel (SNF) vibration integrity was performed in Phase I project under simulated transportation environments, using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot cell testing technology developed at Oak Ridge National Laboratory in 2013–14. The data analysis on the as-irradiated HBR SNF rods demonstrated that the load amplitude is the dominant factor that controls the fatigue life of bending rods. However, previous studies have shown that the hydrogen content and hydride morphology has an important effect on zirconium alloy mechanical properties. To address the effect of radial hydridesmore » in SNF rods, in Phase II a test procedure was developed to simulate the effects of elevated temperatures, pressures, and stresses during transfer-drying operations. Pressurized and sealed fuel segments were heated to the target temperature for a preset hold time and slow-cooled at a controlled rate. The procedure was applied to both non-irradiated/prehydrided and high-burnup Zircaloy-4 fueled cladding segments using the Nuclear Regulatory Commission-recommended 400°C maximum temperature limit at various cooling rates. Before testing high-burnup cladding, four out-of-cell tests were conducted to optimize the hydride reorientation (R) test condition with pre-hydride Zircaloy-4 cladding, which has the same geometry as the high burnup fuel samples. Test HR-HBR#1 was conducted at the maximum hoop stress of 145 MPa, at a 400°C maximum temperature and a 5°C/h cooling rate. On the other hand, thermal cycling was performed for tests HR-HBR#2, HR-HBR#3, and HR-HBR#4 to generate more radial hydrides. It is clear that thermal cycling increases the ratio of the radial hydride to circumferential hydrides. The internal pressure also has a significant effect on the radial hydride morphology. This report describes a procedure and experimental results of the four out-of-cell hydride reorientation tests of hydrided Zircaloy-4 cladding, which served as a guideline to prepare in-cell hydride reorientation samples with high burnup HBR fuel segments. This report also provides the Phase II CIRFT test data for the hydride reorientation irradiated samples. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The CIRFT results appear to indicate that hydride reoriented treatment (HRT) have a negative effect on fatigue life, in addition to hydride reorientation effect. For HR4 specimen that had no pressurization procedure applied, the thermal annealing treatment alone showed a negative impact on the fatigue life compared to the HBR rod.« less

SiC/SiC Cladding Materials Properties Handbook

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

Snead, Mary A.; Katoh, Yutai; Koyanagi, Takaaki

When a new class of material is considered for a nuclear core structure, the in-pile performance is usually assessed based on multi-physics modeling in coordination with experiments. This report aims to provide data for the mechanical and physical properties and environmental resistance of silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites for use in modeling for their application as accidenttolerant fuel cladding for light water reactors (LWRs). The properties are specific for tube geometry, although many properties can be predicted from planar specimen data. This report presents various properties, including mechanical properties, thermal properties, chemical stability under normal and offnormalmore » operation conditions, hermeticity, and irradiation resistance. Table S.1 summarizes those properties mainly for nuclear-grade SiC/SiC composites fabricated via chemical vapor infiltration (CVI). While most of the important properties are available, this work found that data for the in-pile hydrothermal corrosion resistance of SiC materials and for thermal properties of tube materials are lacking for evaluation of SiC-based cladding for LWR applications.« less

Production of FR Tubing from Advanced ODS Alloys

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

Maloy, Stuart Andrew; Lavender, Curt; Omberg, Ron

2016-10-25

Significant research is underway to develop LWR nuclear fuels with improved accident tolerance. One of the leading candidate materials for cladding are the FeCrAl alloys. New alloys produced at ORNL called Gen I and Gen II FeCrAl alloys possess excellent oxidation resistance in steam up to 1400°C and in parallel methods are being developed to produce tubing from these alloys. Century tubing continues to produce excellent tubing from FeCrAl alloys. This memo reports receipt of ~21 feet of Gen I FeCrAl alloy tubing. This tubing will be used for future tests including burst testing, mechanical testing and irradiation testing.

Improved OSC Amtec generator design to meet goals of JPL's candidate Europa Orbiter mission

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

Schock, A.; Noravian, H.; Or, C.

1998-07-01

The preceding paper (Paper IECEC.98.244) described OSC's initial designs of AMTEC (Alkali Metal Thermal-to-Electrical Conversion) power systems, consisting of one or two generators, each with 2, 3, or 4 General Purpose Heat Source (GPHS) modules and with 16 refractory AMTEC cells containing 5 Beta Alumina Solid Electrolyte (BASE) tubes; and presented the effect of heat input and voltage output on the generator's BOM evaporator and clad temperatures and on its EOM system efficiency and power output. Comparison of the computed results with JPL's goals for the Europa Orbiter mission showed that all of the initial 16-cell design options yielded eithermore » excessive evaporator and clad temperatures or insufficient EOM power to satisfy the JPL-specified mission goals. The present paper describes modified OSC generator designs with different numbers of AMTEC cells, cell diameters, cell lengths, cell materials, BASE tube lengths, and number of tubes per cell. These efforts succeeded in identifying generator designs with only half the number of AMTEC cells which -- for the same assumptions -- can produce EOM power outputs substantially in excess of JPL's goals for NASA's Europa Orbiter mission while operating well below the prescribed BOM limits on evaporator and clad temperature; and revealed that lowering the emissivity of the generator's housing to raise the cells' condenser temperatures can achieve substantial additional performance improvement. Finally, the paper culminates in programmatic recommendations.« less

Fuel assembly for the production of tritium in light water reactors

DOEpatents

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

1983-06-10

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

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, William E.; Trapp, Turner J.

1985-01-01

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

Electronic structure calculations of PbS quantum rods and tubes

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

Pimachev, Artem; Dahnovsky, Yuri, E-mail: yurid@uwyo.edu

2014-01-28

We study absorption spectra, optical and HOMO-LUMO gaps, and the density of states for PbS quantum rods (QRs) and tubes (QTs). We find some similarities and also differences in QR and QT properties. For both QRs and QTs, the optical and HOMO-LUMO gaps reach the plateaus for small lengths. We find that tubes are as stable as rods. The optical spectra exhibit a peak that can be due to the electron-hole interaction or be a prototype of an S{sub e}–S{sub h} transition in the effective mass approximation. We also calculate the density of states by the density functional theory (DFT)more » and time-dependent density functional theory (TDDFT) methods. The TDDFT density of states function is shifted towards the red side by 0.5 eV indicating the strong e-h interaction.« less

Expanded plug method for developing circumferential mechanical properties of tubular materials

DOEpatents

Hendrich, William Ray; McAfee, Wallace Jefferson; Luttrell, Claire Roberta

2006-11-28

A method for determining the circumferential properties of a tubular product, especially nuclear fuel cladding, utilizes compression of a polymeric plug within the tubular product to determine strain stress, yield stress and other properties. The process is especially useful in the determination of aging properties such as fuel rod embrittlement after long burn-down.

PROCESS OF FORMING POWDERED MATERIAL

DOEpatents

Glatter, J.; Schaner, B.E.

1961-07-14

A process of forming high-density compacts of a powdered ceramic material is described by agglomerating the powdered ceramic material with a heat- decompossble binder, adding a heat-decompossble lubricant to the agglomerated material, placing a quantity of the material into a die cavity, pressing the material to form a compact, pretreating the compacts in a nonoxidizing atmosphere to remove the binder and lubricant, and sintering the compacts. When this process is used for making nuclear reactor fuel elements, the ceramic material is an oxide powder of a fissionsble material and after forming, the compacts are placed in a cladding tube which is closed at its ends by vapor tight end caps, so that the sintered compacts are held in close contact with each other and with the interior wall of the cladding tube.

Morphology-Dependent Hardness of Cr7C3-Ni-Rich Alloy Composite vs Orientation Independent Hardness of Cr7C3 Primary Phase in a Laser Clad Microstructure

NASA Astrophysics Data System (ADS)

Venkatesh, Lakshmi Narayanan; Suresh Babu, Pitchuka; Gundakaram, Ravi Chandra; Doherty, Roger D.; Joshi, Shrikant V.; Samajdar, Indradev

2017-04-01

Microstructural evolution with superheating was studied in chromium carbide-nickel coatings deposited by laser cladding. At lower superheating, selective growth of <0001> direction from the high density of Cr7C3 grains nucleated resulted in a columnar structure with (0001) texture. Increased superheating lead to the loss of columnar structure as well as the (0001) texture. The hexagonal Cr7C3 showed an unusual isotropic nanoindentation hardness evidently correlated with its low c/ a ratio. However, the rod-like morphology of the carbide dendrites resulted in significant anisotropy in the hardness of the composite.

77 FR 63712 - Airworthiness Directives; Piaggio Aero Industries S.p.A.

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-17

... authority for Italy. We issued that AD to prevent the brake hydraulic fluid from leaking because of the brake assembly rods contacting the brake valve tubing, which could result in the inability to adequately... brake pads between pilot and copilot, a brake system rod was found deflected. The rod, in this bent...

Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report

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

Rosenbaum, H.S.

1979-03-01

This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel, and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress, and reactive fission products during reactor service. This is the final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCImore » far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel > 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2.« less

Torsion testing for general constitutive relations: Gilles Canova's Master's Thesis

NASA Astrophysics Data System (ADS)

Kocks, U. F.; Stout, M. G.

1999-09-01

Torsion testing is useful but cumbersome: useful as a technique to determine large-strain plastic behaviour in a plane-strain mode; cumbersome because the testing of a solid rod provides reliable data in only the simplest of circumstances, for example, when there is no strain hardening or rate sensitivity. The testing of short thin-walled tubes is widely regarded as the best current technique to determine general constitutive behaviour; a drawback is the requirement for large specimen blanks (say, 3 cm cube) and the complex machining procedure. Gilles Canova proposed an alternative - the testing of a series of solid rods of differing diameters, proving the principle by using 10 such rods for one test. We have undertaken a similar series of tests on just four rods plus one short tube for comparison. We found the results from the two types of torsion test in excellent agreement; however, it was not a critical test, inasmuch as the rate sensitivity of the pure copper used was too small. Had it been greater (as, for example, in aluminium and at higher temperature), the evaluation of perhaps six rods would have provided the constitutive response with regard to both hardening and rate sensitivity at the same time - which would require two short-tube tests (plus duplications). The main drawback of the multiple-rod test is that it requires considerable numerical effort in the evaluation. A closer integration with modelling of the constitutive behaviour would be helpful.

Fission gas release during power bumping at high burnup

NASA Astrophysics Data System (ADS)

Cunningham, M. E.; Freshley, M. D.; Lanning, D. D.

1993-03-01

Research to define the behavior of Zircaloy-clad light-water reactor fuel irradiated to high burnup levels was conducted by the High Burnup Effects Program (HBEP). One activity conducted by the HBEP was to "bump" the power level of irradiated, commercial light-water reactor fuel rods to design limit linear heat generation rates at end-of-life. These bumping irradiations simulated end-of-life design limit linear heat generation rates and provided data on the effects of short-term, high power irradiations at high burnup applicable to the design and operating constraints imposed by maximum allowable fuel rod internal gas pressure limits. Based on net fission gas release during the bumping irradiations, it was observed that higher burnup rods had greater rod-average fractional fission gas release than lower burnup rods at equal bumping powers. It was also observed that a hold period of 48 hours at the peak power was insufficient to achieve equilibrium fission gas release. Finally, differences in the prebump location of fission gas, i.e., within the UO 2 matrix or at grain boundaries, affected the fission gas release during the bumping irradiations.

METHOD FOR MAKING FUEL ELEMENTS

DOEpatents

Kates, L.W.; Campbell, R.W.; Heartel, R.H.W.

1960-08-01

A method is given for making zirconium-clad uranium wire. A tube of zirconium is closed with a zirconium plug, after which a chilled uranium core is inserted in the tube to rest against the plug. Additional plugs and cores are inserted alternately as desired. The assembly is then sheathed with iron, hot worked to the desired size, and the iron sheath removed.

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.

Tensile properties from room temperature to 1315 C of tungsten-lined tantalum-alloy (T-111) tubing fabricated by hot isostatic pressing

NASA Technical Reports Server (NTRS)

Buzzard, R. J.; Metroka, R. R.

1974-01-01

The effects were studied of a thin tungsten liner on the tensile properties of T-111 tubing considered for fuel cladding in a space power nuclear reactor concept. The results indicate that the metallurgically bonded liner had no appreciable effects on the properties of the T-111 tubing. A hot isostatic pressing method used to apply the liners is described along with a means for overcoming the possible embrittling effects of hydrogen contamination.

Tubing and cable cutting tool

NASA Technical Reports Server (NTRS)

Mcsmith, D. D.; Richardson, J. I. (Inventor)

1984-01-01

A hand held hydraulic cutting tool was developed which is particularly useful in deactivating ejection seats in military aircraft rescue operations. The tool consists primarily of a hydraulic system composed of a fluid reservoir, a pumping piston, and an actuator piston. Mechanical cutting jaws are attached to the actuator piston rod. The hydraulic system is controlled by a pump handle. As the pump handle is operated the actuator piston rod is forced outward and thus the cutting jaws are forced together. The frame of the device is a flexible metal tubing which permits easy positioning of the tool cutting jaws in remote and normally inaccessible locations. Bifurcated cutting edges ensure removal of a section of the tubing or cable to thereby reduce the possibility of accidental reactivation of the tubing or cable being severed.

History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

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

Larry Zirker; Nathan Jerred; Dr. Indrajit Charit

2012-03-01

Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

Investigation of Zircaloy-2 oxidation model for SFP accident analysis

NASA Astrophysics Data System (ADS)

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu; Nakashima, Kazuo; Kanazawa, Toru; Tojo, Masayuki

2017-05-01

The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study.

Further observations on OCOM MOX fuel: microstructure in the vicinity of the pellet rim and fuel — cladding interaction

NASA Astrophysics Data System (ADS)

Walker, C. T.; Goll, W.; Matsumura, T.

1997-06-01

The fuel investigated was manufactured by Siemens-KWU and irradiated at low rating in the KWO reactor in Germany. The MOX agglomerates in the cold outer region of the fuel shared several common features with the high burn-up structure at the rim of UO 2 fuel. It is proposed that in both cases the mechanism producing the microstructure change is recrystallisation. Further, it is shown that surface MOX agglomerates do not noticeably retard cladding creepdown although they swell into the gap. The contracting cladding appears able to push the agglomerates back into the fuel. The thickness of the oxide layer on the inner cladding surface increased at points where contact with surface MOX agglomerates had occurred. Despite this, the mean thickness of the oxide did not differ significantly from that found in UO 2 fuel rods of like design. It is judged that the high burn-up structure will form in the UO 2 matrix when the local burn-up there reaches 60 to 80 GWd/tM. Limiting the MOX scrap addition in the UO 2 matrix will delay its formation.

Benchmarking the SPHINX and CTH shock physics codes for three problems in ballistics

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

Wilson, L.T.; Hertel, E.; Schwalbe, L.

1998-02-01

The CTH Eulerian hydrocode, and the SPHINX smooth particle hydrodynamics (SPH) code were used to model a shock tube, two long rod penetrations into semi-infinite steel targets, and a long rod penetration into a spaced plate array. The results were then compared to experimental data. Both SPHINX and CTH modeled the one-dimensional shock tube problem well. Both codes did a reasonable job in modeling the outcome of the axisymmetric rod impact problem. Neither code correctly reproduced the depth of penetration in both experiments. In the 3-D problem, both codes reasonably replicated the penetration of the rod through the first plate.more » After this, however, the predictions of both codes began to diverge from the results seen in the experiment. In terms of computer resources, the run times are problem dependent, and are discussed in the text.« less

Use of Site Characterization and Analysis Penetrometer System at the Walnut Creek Watershed, Ames, Iowa

DTIC Science & Technology

1993-08-01

grout is pumped through a central tube in the push rod cable and exits through the probe tip. Retraction grouting (as the rod is withdrawn through the...soil) with a microfine Portland cement grout is normally utilized for all projects; however, subfreezing weather mandated that post - retraction grouting...ELECTRODES (4) P2 RESISTIVITY C2 MODULE GROUTING TUBES (2) 24 IN(61 CM) Grout Injector -- 1.4 IN(3.57 CM) o.D. 0" RINGS GROUTING CONE SLEEVE FRICTION

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

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

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)« less

High burnup fuel behavior related to fission gas effects under reactivity initiated accidents (RIA) conditions

NASA Astrophysics Data System (ADS)

Lemoine, F.

1997-09-01

Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning the transient fission gas behavior and the clad loading mechanisms. The importance of the rim zone is demonstrated based on three experiments resulting in clad failure at low enthalpy, which are explained by energetic considerations. High gas release in non-failure tests with low energy deposition underlines the importance of grain boundary and porosity gas. Measured final releases are strongly correlated to the microstructure evolution, depending on energy deposition, pulse width, initial and refabricated fuel rod design. Observed helium release can also increase internal pressure and gives hints to the gas behavior understanding.

Development of a (147)Pm source for beta-backscatter thickness gauge applications.

PubMed

Kumar, Manoj; Udhayakumar, J; Nuwad, J; Shukla, Rakesh; Pillai, C G S; Dash, Ashutosh; Venkatesh, Meera

2011-03-01

This paper describes a method for the preparation of (147)Pm sources, utilized in the determination of graphite coating thickness on the inner surface of the zircaloy cladding tube of nuclear fuels. (147)Pm was adsorbed on a limited surface area [1.5mm (ϕ)×2mm (l)] of a cylindrical aluminum rod [1.5mm (ϕ)×10mm (l)]. In brief, the selected tip area [1.5mm (ϕ)×2mm (l)] was anodized at a current density of 15mA/cm(2) at 15°C in 3M·H(2)SO(4) for 2h followed by immersion of this area in 10μL of (147)Pm solution containing 37MBq (1mCi) of activity at pH 6.0 for 24h. The radioactive area was subsequently covered with a thin layer of Polymethyl Methacrylate (PMMA) to prevent leaching of (147)Pm from the source. The quantity of incorporated (147)Pm activity was assayed in a calibrated ion chamber. Quality control tests were carried out to ensure nonleachability, uniform distribution of activity and stability of the sources. Copyright © 2010 Elsevier Ltd. All rights reserved.

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

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

NASA Astrophysics Data System (ADS)

Aydogan, E.; Weaver, J. S.; Maloy, S. A.; El-Atwani, O.; Wang, Y. Q.; Mara, N. A.

2018-05-01

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al2O3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe2+ ion irradiation up to ∼16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two-beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size and a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α‧ precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ∼3.4 dpa and ∼16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

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

Aydogan, E.; Weaver, J. S.; Maloy, S. A.

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al 2O 3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe 2+ ion irradiation up to ~16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size andmore » a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α' precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ~3.4 dpa and ~16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.« less

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

DOE PAGES

Aydogan, E.; Weaver, J. S.; Maloy, S. A.; ...

2018-03-02

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al 2O 3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe 2+ ion irradiation up to ~16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size andmore » a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α' precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ~3.4 dpa and ~16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.« less

Microfabricated bragg waveguide

DOEpatents

Fleming, James G.; Lin, Shawn-Yu; Hadley, G. Ronald

2004-10-19

A microfabricated Bragg waveguide of semiconductor-compatible material having a hollow core and a multilayer dielectric cladding can be fabricated by integrated circuit technologies. The microfabricated Bragg waveguide can comprise a hollow channel waveguide or a hollow fiber. The Bragg fiber can be fabricated by coating a sacrificial mandrel or mold with alternating layers of high- and low-refractive-index dielectric materials and then removing the mandrel or mold to leave a hollow tube with a multilayer dielectric cladding. The Bragg channel waveguide can be fabricated by forming a trench embedded in a substrate and coating the inner wall of the trench with a multilayer dielectric cladding. The thicknesses of the alternating layers can be selected to satisfy the condition for minimum radiation loss of the guided wave.

Development of Cone Wedge Ring Expansion Test to Evaluate Mechanical Properties of Clad Tubing Structure

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

Wang, Jy-An John

To determine the hoop tensile properties of irradiated fuel cladding in a hot cell, a cone wedge ring expansion test method was developed. A four-piece wedge insert was designed with tapered angles matched to the cone shape of a loading piston. The ring specimen was expanded in the radial direction by the lateral expansion of the wedges under the downward movement of the piston. The advantages of the proposed method are that implementation of the test setup in a hot cell is simple and easy, and that it enables a direct strain measurement of the test specimen from the piston’smore » vertical displacement soon after the wedge-clad contact resistance is initiated.« less

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

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

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

Development of single crystal membranes

NASA Technical Reports Server (NTRS)

Stormont, R. W.; Cocks, F. H.

1972-01-01

The design and construction of a high pressure crystal growth chamber was accomplished which would allow the growth of crystals under inert gas pressures of 2 MN/sq m (300 psi). A novel crystal growth technique called EFG was used to grow tubes and rods of the hollandite compounds, BaMgTi7O16, K2MgTi7O16, and tubes of sodium beta-alumina, sodium magnesium-alumina, and potassium beta-alumina. Rods and tubes grown are characterized using metallographic and X-ray diffraction techniques. The hollandite compounds are found to be two or three-phase, composed of coarse grained orientated crystallites. Single crystal c-axis tubes of sodium beta-alumina were grown from melts containing excess sodium oxide. Additional experiments demonstrated that crystals of magnesia doped beta-alumina and potassium beta-alumina also can be achieved by this EFG technique.

CONTROL RODS FOR NUCLEAR REACTORS

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

Bell, F.R.

1963-01-16

A means for controlling the control rod in emergency, when it is desired to shutdown the reactor with the shortest possible delay, is described. When the emergency occurs the control rod is allowed to drop freely under gravity from the control rod support tube into the bore in the reactor core. A normal shutdown is reached almost at the lowest rod position. In the shut-down position and also below it, the control rod had its full effect of reducing the level of activity in the core. When the shut-down position was reached, a brake came into action to decelerate themore » rod and reduce shock and the likelihood of damage. (C.E.S.)« less

Comparisons of Wilks’ and Monte Carlo Methods in Response to the 10CFR50.46(c) Proposed Rulemaking

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

Zhang, Hongbin; Szilard, Ronaldo; Zou, Ling

The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10CFR50.46(c), the US NRC put forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently nearly all the fuel rods in a reactor core need to be analyzed under LOCA conditions to demonstrate compliance to the safety limits. New analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of themore » limiting rods as well as to quantify the safety margins under LOCA conditions. With the new analysis method presented in this work, the limiting transient case and the limiting rods can be easily identified to quantify the safety margins in response to the proposed new rulemaking. In this work, the best-estimate plus uncertainty (BEPU) analysis capability for large break LOCA with the new cladding embrittlement criteria using the RELAP5-3D code is established and demonstrated with a reduced set of uncertainty parameters. Both the direct Monte Carlo method and the Wilks’ nonparametric statistical method can be used to perform uncertainty quantification. Wilks’ method has become the de-facto industry standard to perform uncertainty quantification in BEPU LOCA analyses. Despite its widespread adoption by the industry, the use of small sample sizes to infer statement of compliance to the existing 10CFR50.46 rule, has been a major cause of unrealized operational margin in today’s BEPU methods. Moreover the debate on the proper interpretation of the Wilks’ theorem in the context of safety analyses is not fully resolved yet, even more than two decades after its introduction in the frame of safety analyses in the nuclear industry. This represents both a regulatory and application risk in rolling out new methods. With the 10CFR50.46(c) proposed rulemaking, the deficiencies of the Wilks’ approach are further exacerbated. The direct Monte Carlo approach offers a robust alternative to perform uncertainty quantification within the context of BEPU analyses. In this work, the Monte Carlo method is compared with the Wilks’ method in response to the NRC 10CFR50.46(c) proposed rulemaking.« less

Statistical analysis of QC data and estimation of fuel rod behaviour

NASA Astrophysics Data System (ADS)

Heins, L.; Groβ, H.; Nissen, K.; Wunderlich, F.

1991-02-01

The behaviour of fuel rods while in reactor is influenced by many parameters. As far as fabrication is concerned, fuel pellet diameter and density, and inner cladding diameter are important examples. Statistical analyses of quality control data show a scatter of these parameters within the specified tolerances. At present it is common practice to use a combination of superimposed unfavorable tolerance limits (worst case dataset) in fuel rod design calculations. Distributions are not considered. The results obtained in this way are very conservative but the degree of conservatism is difficult to quantify. Probabilistic calculations based on distributions allow the replacement of the worst case dataset by a dataset leading to results with known, defined conservatism. This is achieved by response surface methods and Monte Carlo calculations on the basis of statistical distributions of the important input parameters. The procedure is illustrated by means of two examples.

Partially-reflected water-moderated square-piteched U(6.90)O 2 fuel rod lattices with 0.67 fuel to water volume ratio (0.800 CM Pitch)

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

Harms, Gary A.

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% 235U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O 2 fuel rods.

Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit

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

John H. Strumpell

2004-12-31

Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and/or disposal and enable plants to operate with longer more economical cycle lengths and/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding material and have irradiated several test fuel rods through four cycles. The U.S. Department of Energy (DOE) joined FANP and the BWOG in supporting this project during its final phase of collecting and evaluating high burnup data through post irradiationmore » examination (PIE).« less

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

Singer, W.; Singer, X.; Jelezov, I.

Activities of the past several years in developing the technique of forming seamless (weldless) cavity cells by hydroforming are summarized. An overview of the technique developed at DESY for the fabrication of single cells and multicells of the TESLA cavity shape is given and the major rf results are presented. The forming is performed by expanding a seamless tube with internal water pressure while simultaneously swaging it axially. Prior to the expansion the tube is necked at the iris area and at the ends. Tube radii and axial displacements are computer controlled during the forming process in accordance with resultsmore » of finite element method simulations for necking and expansion using the experimentally obtained strain-stress relationship of tube material. In cooperation with industry different methods of niobium seamless tube production have been explored. The most appropriate and successful method is a combination of spinning or deep drawing with flow forming. Several single-cell niobium cavities of the 1.3 GHz TESLA shape were produced by hydroforming. They reached accelerating gradients E acc up to 35 MV/m after buffered chemical polishing (BCP) and up to 42 MV/m after electropolishing (EP). More recent work concentrated on fabrication and testing of multicell and nine-cell cavities. Several seamless two- and three-cell units were explored. Accelerating gradients E acc of 30–35 MV/m were measured after BCP and E acc up to 40 MV/m were reached after EP. Nine-cell niobium cavities combining three three-cell units were completed at the company E. Zanon. These cavities reached accelerating gradients of E acc = 30–35 MV/m. One cavity is successfully integrated in an XFEL cryomodule and is used in the operation of the FLASH linear accelerator at DESY. Additionally the fabrication of bimetallic single-cell and multicell NbCu cavities by hydroforming was successfully developed. Several NbCu clad single-cell and double-cell cavities of the TESLA shape have been fabricated. The clad seamless tubes were produced using hot bonding or explosive bonding and subsequent flow forming. The thicknesses of Nb and Cu layers in the tube wall are about 1 and 3 mm respectively. The rf performance of the best NbCu clad cavities is similar to that of bulk Nb cavities. The highest accelerating gradient achieved was 40 MV/m. The advantages and disadvantages of hydroformed cavities are discussed in this paper.« less

Hydroforming of elliptical cavities

DOE PAGES

Singer, W.; Singer, X.; Jelezov, I.; ...

2015-02-27

Activities of the past several years in developing the technique of forming seamless (weldless) cavity cells by hydroforming are summarized. An overview of the technique developed at DESY for the fabrication of single cells and multicells of the TESLA cavity shape is given and the major rf results are presented. The forming is performed by expanding a seamless tube with internal water pressure while simultaneously swaging it axially. Prior to the expansion the tube is necked at the iris area and at the ends. Tube radii and axial displacements are computer controlled during the forming process in accordance with resultsmore » of finite element method simulations for necking and expansion using the experimentally obtained strain-stress relationship of tube material. In cooperation with industry different methods of niobium seamless tube production have been explored. The most appropriate and successful method is a combination of spinning or deep drawing with flow forming. Several single-cell niobium cavities of the 1.3 GHz TESLA shape were produced by hydroforming. They reached accelerating gradients E acc up to 35 MV/m after buffered chemical polishing (BCP) and up to 42 MV/m after electropolishing (EP). More recent work concentrated on fabrication and testing of multicell and nine-cell cavities. Several seamless two- and three-cell units were explored. Accelerating gradients E acc of 30–35 MV/m were measured after BCP and E acc up to 40 MV/m were reached after EP. Nine-cell niobium cavities combining three three-cell units were completed at the company E. Zanon. These cavities reached accelerating gradients of E acc = 30–35 MV/m. One cavity is successfully integrated in an XFEL cryomodule and is used in the operation of the FLASH linear accelerator at DESY. Additionally the fabrication of bimetallic single-cell and multicell NbCu cavities by hydroforming was successfully developed. Several NbCu clad single-cell and double-cell cavities of the TESLA shape have been fabricated. The clad seamless tubes were produced using hot bonding or explosive bonding and subsequent flow forming. The thicknesses of Nb and Cu layers in the tube wall are about 1 and 3 mm respectively. The rf performance of the best NbCu clad cavities is similar to that of bulk Nb cavities. The highest accelerating gradient achieved was 40 MV/m. The advantages and disadvantages of hydroformed cavities are discussed in this paper.« less

Hydroforming of elliptical cavities

NASA Astrophysics Data System (ADS)

Singer, W.; Singer, X.; Jelezov, I.; Kneisel, P.

2015-02-01

Activities of the past several years in developing the technique of forming seamless (weldless) cavity cells by hydroforming are summarized. An overview of the technique developed at DESY for the fabrication of single cells and multicells of the TESLA cavity shape is given and the major rf results are presented. The forming is performed by expanding a seamless tube with internal water pressure while simultaneously swaging it axially. Prior to the expansion the tube is necked at the iris area and at the ends. Tube radii and axial displacements are computer controlled during the forming process in accordance with results of finite element method simulations for necking and expansion using the experimentally obtained strain-stress relationship of tube material. In cooperation with industry different methods of niobium seamless tube production have been explored. The most appropriate and successful method is a combination of spinning or deep drawing with flow forming. Several single-cell niobium cavities of the 1.3 GHz TESLA shape were produced by hydroforming. They reached accelerating gradients Eacc up to 35 MV /m after buffered chemical polishing (BCP) and up to 42 MV /m after electropolishing (EP). More recent work concentrated on fabrication and testing of multicell and nine-cell cavities. Several seamless two- and three-cell units were explored. Accelerating gradients Eacc of 30 - 35 MV /m were measured after BCP and Eacc up to 40 MV /m were reached after EP. Nine-cell niobium cavities combining three three-cell units were completed at the company E. Zanon. These cavities reached accelerating gradients of Eacc=30 - 35 MV /m . One cavity is successfully integrated in an XFEL cryomodule and is used in the operation of the FLASH linear accelerator at DESY. Additionally the fabrication of bimetallic single-cell and multicell NbCu cavities by hydroforming was successfully developed. Several NbCu clad single-cell and double-cell cavities of the TESLA shape have been fabricated. The clad seamless tubes were produced using hot bonding or explosive bonding and subsequent flow forming. The thicknesses of Nb and Cu layers in the tube wall are about 1 and 3 mm respectively. The rf performance of the best NbCu clad cavities is similar to that of bulk Nb cavities. The highest accelerating gradient achieved was 40 MV /m . The advantages and disadvantages of hydroformed cavities are discussed in this paper.

Corrosion and hydrogen pick-up behaviors of cladding and structural components in BWR high burnup 9x9 lead use assemblies

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

Miyashita, Toshiyasu; Nakae, Nobuo; Ogata, Keizo

The high burnup BWR 9x9 lead use fuel assemblies, which have been designed for maximum assembly burnup of 55 GWd/t in Japan, have been examined after irradiations to confirm the reliability of the current safety evaluation methodology, and to accumulate data to judge the adequacy to apply it to the future higher burnup fuel. After 3 and 5 cycle irradiations, post irradiation examinations were performed for both 9x9 Type-A and Type-B fuel assemblies. Both Type LUAs utilize Zry-2 claddings, while there are deviation in the contents of impurity and alloying elements between Type-A and Type-B, especially in Fe and Simore » concentration. Measured oxide thicknesses of fuel rods showed no significant difference between after 3 and 5 cycle irradiation except for some rods at corner position in Type B LUA. The axial profile of hydrogen concentration and oxide thickness for the corner rods in Type B LUA after 5 cycle irradiation had peaks at the second lowest span from the bottom. The maximum oxide thickness is about 50 {mu}m on the surface facing the bundle outside at the second lowest span and dense hydrides layer (Hydride rim) is observed in peripheral region of cladding showing unexpected high hydrogen concentration. The results of calculated thermal-hydraulic conditions show that the thermal neutron flux at the corner position was higher than the other position. On the other hand, the void fraction and the mass flux were relatively lower at the corner position. The oxide thickness on spacer band and spacer cell of Zry-2 increases from 3 to 5 cycle irradiations. Spacer band of Zry-4 showed significantly thick oxide after 5 cycle irradiations but Hydrogen concentration was relatively small in contrast its obviously thick oxide in comparison with Zry-2 spacer bands. The large increase in hydrogen concentration was measured in Zry-2 spacers after 5 cycle irradiations and the evaluated hydrogen pick-up rate also increased remarkably. (authors)« less

Guiding and amplification properties of rod-type photonic crystal fibers with sectioned core doping

NASA Astrophysics Data System (ADS)

Selleri, S.; Poli, F.; Passaro, D.; Cucinotta, A.; Lægsgaard, J.; Broeng, J.

2009-05-01

Rod-type photonic crystal fibers are large mode area double-cladding fibers with an outer diameter of few millimeters which can provide important advantages for high-power lasers and amplifiers. Numerical studies have recently demonstrated the guidance of higher-order modes in these fibers, which can worsen the output beam quality of lasers and amplifiers. In the present analysis a sectioned core doping has been proposed for Ybdoped rod-type photonic crystal fibers, with the aim to improve the higher-order mode suppression. A full-vector modal solver based on the finite element method has been applied to properly design the low refractive index ring in the fiber core, which can provide an increase of the differential overlap between the fundamental and the higher-order mode. Then, the gain competition among the guided modes along the Yb-doped rod-type fibers has been investigated with a spatial and spectral amplifier model. Simulation results have shown the effectiveness of the sectioned core doping in worsening the higher-order mode overlap on the doped area, thus providing an effective single-mode behavior of the Yb-doped rod-type photonic crystal fibers.

Heat transfer and pressure drop studies of TiO2/DI water nanofluids in helically corrugated tubes using spiraled rod inserts

NASA Astrophysics Data System (ADS)

Anbu, S.; Venkatachalapathy, S.; Suresh, S.

2018-05-01

An experimental study on the convective heat transfer and friction factor characteristics of TiO2/DI water nanofluids in uniformly heated plain and helically corrugated tubes (HCT) with and without spiraled rod inserts (SRI) under laminar flow regime is presented in this paper. TiO2 nanoparticles with an average size of 32 nm are dispersed in deionized (DI) water to form stable suspensions containing 0.1, 0.15, 0.2, and 0.25% volume concentrations of nanoparticles. It is found that the inclusion of nanoparticles to DI water ameliorated Nusselt number which increased with nanoparticles concentration upto 0.2%. Two spiraled rod inserts made of copper with different pitches (pi = 50 mm and 30 mm) are inserted in both plain and corrugated tubes and it is found that the addition of these inserts increased the Nusselt number substantially. For Helically corrugated tube with lower pitch and maximum height of corrugation (pc = 8 mm, hc = 1 mm) with 0.2% volume concentration of nanoparticles, a maximum enhancement of 15% in Nusselt number is found without insert and with insert having lower pitch (pi = 30 mm) the enhancement is 34% when compared to DI water in plain tube. The results on friction factor show a maximum penalty of about 53.56% for the above HCT.

77 FR 74170 - Foreign-Trade Zone 84-Houston, TX; Notification of Proposed Production Activity; Mitsubishi...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-13

..., tubes, pipes, hoses, tires, rods, containers, handles, belts), pallets, wood packing boxes, cases, other..., wrenches, hand tools, flexible tubing, engines, parts of engines, water boilers, control panels, control...

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

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.

Ophthalmic method and apparatus

NASA Technical Reports Server (NTRS)

Evvard, J. C.; Mcgannon, W. J.; Vargo, D. J. (Inventor)

1973-01-01

A method and apparatus for removing material and components such as the lens from an eye is described. High speed rotary cutting members at one end of a rod macerate the lens while an annular tubing disposed around the cutting members vibrates ultrasonically to coact with the cutting members in macerating the lens. At the same time, a liquid is supplied to the chamber behind the cornea of the eye. Spiral grooves extending along the rotating rod from the cutting members evacuate the liquid and the macerated material from the eye. An alternate embodiment of the apparatus includes a tube through which liquid is supplied to the operative site of the ultrasonically vibrating tube and the cutting members in the area of the lens.

Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

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

Szilard, Ronaldo Henriques; Youngblood, Robert; Frepoli, Cesare

2015-04-01

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review andmore » approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.« less

Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

NASA Astrophysics Data System (ADS)

Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

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

Szilard, Ronaldo H.; Youngblood, Robert W.; Zhang, Hongbin

2015-09-01

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions formore » NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.« less

Wastewater heat recovery method and apparatus

DOEpatents

Kronberg, J.W.

1991-01-01

This invention is comprised of a heat recovery system with a heat exchanger and a mixing valve. A drain trap includes a heat exchanger with an inner coiled tube, baffle plate, wastewater inlet, wastewater outlet, cold water inlet, and preheated water outlet. Wastewater enters the drain trap through the wastewater inlet, is slowed and spread by the baffle plate, and passes downward to the wastewater outlet. Cold water enters the inner tube through the cold water inlet and flows generally upward, taking on heat from the wastewater. This preheated water is fed to the mixing valve, which includes a flexible yoke to which are attached an adjustable steel rod, two stationary zinc rods, and a pivoting arm. The free end of the arm forms a pad which rests against a valve seat. The rods and pivoting arm expand or contract as the temperature of the incoming preheated water changes. The zinc rods expand more than the steel rod, flexing the yoke and rotating the pivoting arm. The pad moves towards the valve seat as the temperature of the preheated water rises, and away as the temperature falls, admitting a variable amount of hot water to maintain a nearly constant average process water temperature.

Wastewater heat recovery apparatus

DOEpatents

Kronberg, James W.

1992-01-01

A heat recovery system with a heat exchanger and a mixing valve. A drain trap includes a heat exchanger with an inner coiled tube, baffle plate, wastewater inlet, wastewater outlet, cold water inlet, and preheated water outlet. Wastewater enters the drain trap through the wastewater inlet, is slowed and spread by the baffle plate, and passes downward to the wastewater outlet. Cold water enters the inner tube through the cold water inlet and flows generally upward, taking on heat from the wastewater. This preheated water is fed to the mixing valve, which includes a flexible yoke to which are attached an adjustable steel rod, two stationary zinc rods, and a pivoting arm. The free end of the arm forms a pad which rests against a valve seat. The rods and pivoting arm expand or contract as the temperature of the incoming preheated water changes. The zinc rods expand more than the steel rod, flexing the yoke and rotating the pivoting arm. The pad moves towards the valve seat as the temperature of the preheated water rises, and away as the temperature falls, admitting a variable amount of hot water to maintain a nearly constant average process water temperature.

Wastewater heat recovery apparatus

DOEpatents

Kronberg, J.W.

1992-09-01

A heat recovery system is described with a heat exchanger and a mixing valve. A drain trap includes a heat exchanger with an inner coiled tube, baffle plate, wastewater inlet, wastewater outlet, cold water inlet, and preheated water outlet. Wastewater enters the drain trap through the wastewater inlet, is slowed and spread by the baffle plate, and passes downward to the wastewater outlet. Cold water enters the inner tube through the cold water inlet and flows generally upward, taking on heat from the wastewater. This preheated water is fed to the mixing valve, which includes a flexible yoke to which are attached an adjustable steel rod, two stationary zinc rods, and a pivoting arm. The free end of the arm forms a pad which rests against a valve seat. The rods and pivoting arm expand or contract as the temperature of the incoming preheated water changes. The zinc rods expand more than the steel rod, flexing the yoke and rotating the pivoting arm. The pad moves towards the valve seat as the temperature of the preheated water rises, and away as the temperature falls, admitting a variable amount of hot water to maintain a nearly constant average process water temperature. 6 figs.

Initial experimental evaluation of crud-resistant materials for light water reactors

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Carlson, M. B.; Hussey, D.; Short, M. P.

2018-01-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues.

High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

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

Perez, Emmanuel; Keiser, Jr., Dennis D.; Forsmann, Bryan

High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or betweenmore » the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.« less

Fail-safe storage rack for irradiated fuel rod assemblies

DOEpatents

Lewis, D.R.

1993-03-23

A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

Fail-safe storage rack for irradiated fuel rod assemblies

DOEpatents

Lewis, Donald R.

1993-01-01

A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.; Johnson, H.W.

1961-04-01

BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

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

Montgomery, Robert O.; Bennett, Wendy D.; Henager, Charles H.

2016-09-12

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impactmore » of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.« less

Sapphire tube pressure vessel

DOEpatents

Outwater, John O.

2000-01-01

A pressure vessel is provided for observing corrosive fluids at high temperatures and pressures. A transparent Teflon bag contains the corrosive fluid and provides an inert barrier. The Teflon bag is placed within a sapphire tube, which forms a pressure boundary. The tube is received within a pipe including a viewing window. The combination of the Teflon bag, sapphire tube and pipe provides a strong and inert pressure vessel. In an alternative embodiment, tie rods connect together compression fittings at opposite ends of the sapphire tube.

Stressed waveguides with tubular depressed-cladding inscribed in phosphate glasses by femtosecond hollow laser beams.

PubMed

Long, Xuewen; Bai, Jing; Zhao, Wei; Stoian, Razvan; Hui, Rongqing; Cheng, Guanghua

2012-08-01

We report on the single-step fabrication of stressed optical waveguides with tubular depressed-refractive-index cladding in phosphate glasses by the use of focused femtosecond hollow laser beams. Tubelike low index regions appear under direct exposure due to material rarefaction following expansion. Strained compacted zones emerged in domains neighboring the tubular track of lower refractive index, and waveguiding occurs mainly within the tube core fabricated by the engineered femtosecond laser beam. The refractive index profile of the optical waveguide was reconstructed from the measured transmitted near-field intensity.

Power-Tool Adapter For T-Handle Screws

NASA Technical Reports Server (NTRS)

Deloach, Stephen R.

1992-01-01

Proposed adapter enables use of pneumatic drill, electric drill, electric screwdriver, or similar power tool to tighten or loosen T-handled screws. Notched tube with perpendicular rod welded to it inserted in chuck of tool. Notched end of tube slipped over screw handle.

Lathe Attachment Finishes Inner Surface of Tubes

NASA Technical Reports Server (NTRS)

Lancki, A. J.

1982-01-01

Extremely smooth finishes are machined on inside surfaces of tubes by new attachment for a lathe. The relatively inexpensive accessory, called a "microhone," holds a honing stone against workpiece by rigid tangs instead of springs as in conventional honing tools. Inner rod permits adjustment of microhoning stone, while outer tube supports assembly. Outer tube is held between split blocks on lathe toolpost. Microhoning can be done with either microhone or workpiece moving and other member stationary.

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

Material test machine for tension-compression tests at high temperature

DOEpatents

Cioletti, Olisse C.

1988-01-01

Apparatus providing a device for testing the properties of material specimens at high temperatures and pressures in controlled water chemistries includes, inter alia, an autoclave housing the specimen which is being tested. The specimen is connected to a pull rod which couples out of the autoclave to an external assembly which includes one or more transducers, a force balance chamber and a piston type actuator. The pull rod feeds through the force balance chamber and is compensated thereby for the pressure conditions existing within the autoclave and tending to eject the pull rod therefrom. The upper end of the push rod is connected to the actuator through elements containing a transducer comprising a linear variable differential transformer (LVDT). The housing and coil assembly of the LVDT is coupled to a tube which runs through a central bore of the pull rod into the autoclave where it is connected to one side of the specimen. The movable core of the LVDT is coupled to a stem which runs through the tube where it is then connected to the other side of the specimen through a coupling member. A transducer in the form of a load cell including one or more strain gages is located on a necked-down portion of the upper part of the pull rod intermediate the LVDT and force balance chamber.

Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

DOEpatents

Mariani, Robert Dominick

2014-09-09

Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

High temperature integrated ultrasonic shear and longitudinal wave probes

NASA Astrophysics Data System (ADS)

Ono, Y.; Jen, C.-K.; Kobayashi, M.

2007-02-01

Integrated ultrasonic shear wave probes have been designed and developed using a mode conversion theory for nondestructive testing and characterization at elevated temperatures. The probes consisted of metallic substrates and high temperature piezoelectric thick (>40μm) films through a paint-on method. Shear waves are generated due to mode conversion from longitudinal to shear waves because of reflection inside the substrate having a specific shape. A novel design scheme is proposed to reduce the machining time of substrates and thick film fabrication difficulty. A probe simultaneously generating and receiving both longitudinal and shear waves is also developed and demonstrated. In addition, a shear wave probe using a clad buffer rod consisting of an aluminum core and stainless steel cladding has been developed. All the probes were tested and successfully operated at 150°C.

PM-1 NUCLEAR POWER PLANT PROGRAM. Quarterly Progress Report No. 2 for June 1 to August 31, 1959

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

Sieg, J.S.; Smith, E.H.

1959-10-01

The objective of the contract is the design, development, fabrication, installation, and initial testing and operation of a prepackaged air- transportable pressurized water reactor nuclear power plant, the PM-1. The specified output is 1 Mwe and 7 million Btu/hr of heat. The plant is to be operational by March 1962. The principal efforts were completion of the plant parametric study and preparation of the preliminary design. A summary of design parameters is given. Systems development work included study and selection of packages for full-scale testing, a survey of in-core instrumentation techniques, control and instrumentation development, and development of components formore » the steam generator, condenser, and turbine generator, which are not commercially available. Reactor development work included completion of the parametric zeropower experiments and preparrtions for a flexible zeropower test program, a revision of plans for irradiation testing PM-1 fuel elements, initiation of a reactor flow test program, outliring of a heat tnansfer test program, completion of the seven-tube test section (SETCH-1) tests, and evaluation of control rod actuators leading to specification of a magnetic jack-type control rod drive similar to that reported in ANL-5768. Completion of the prelimirary design led to initiation of the final design effort, which will be the principal activity during the next two project quarters. Preparations for core fabrication included procurement of core cladding material for the zero-power teat core, arrangement with a subcontractor to convent UF/sub 6/ to UO/sub 2/ and to commence delivery of the oxide during the next quarter, development of fuel element fabrication and ultrasonic testing techniques, study of control rod materials, UO/sub 2/ recovery techniques, and boron analysis methods. Preliminary work on site preparation was pursued with receipt of USAEC approval for a location on the eastern slope of Warren Peak at Sundance, Wyoming. A survey of this site is underway. A preliminary Hazards Summary Report is in preparation. (For preceding period see MND-M-1812.) (auth)« less

System for fuel rod removal from a reactor module

DOEpatents

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

1988-07-28

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

System for fuel rod removal from a reactor module

DOEpatents

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

1990-01-01

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

Data summary report for fission product release Test VI-7

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

Osborne, M.F.; Lorentz, R.A.; Travis, J.R.

Test VI-7 was the final test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the Monticello boiling water reactor (BWR). The fuel had experienced a burnup of {approximately}-40 Mwd/kg U. It was heated in an induction furnace for successive 20-min periods at 2000 and 2300 K in a moist air-helium atmosphere. Integral releases were 69% for {sup 85}Kr, 52% for {sup 125}Sb, 71% for both {sup 134}Cs and {sup 137}Cs, and 0.04% for {sup 154}Eu. For the non-gamma-emitting species, release values for 42% for I, 4.1% formore » Ba, 5.3% for Mo, and 1.2% for Sr were determined. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.89 g, with 37% being collected on the thermal gradient tubes and 63% downstream on filters. Posttest examination of the fuel specimen indicated that most of the cladding was completely oxidized to ZrO{sub 2}, but that oxidation was not quite complete at the upper end. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL-Booth Model.« less

4.5 W supercontinuum generation from 1017 to 3438 nm in an all-solid fluorotellurite fiber

NASA Astrophysics Data System (ADS)

Jia, Zhixu; Yao, Chuanfei; Jia, Shijie; Wang, Fang; Wang, Shunbin; Zhao, Zhipeng; Qin, Guanshi; Ohishi, Yasutake; Qin, Weiping

2017-06-01

All-solid fluorotellurite fibers are fabricated by using a rod-in-tube method. The core and cladding materials are TeO2-BaF2-Y2O3 (TBY) and AlF3-based glasses, respectively. Since the refractive index (˜1.46) of AlF3-based glass is much lower than that (˜1.84) of TBY glass, the zero-dispersion-wavelength of the fabricated fiber can be tuned from 2145 to 1507 nm by varying the fiber core diameter from 50 to 3 μm. By using a 0.6 m long all-solid fluorotellurite fiber with a core diameter of ˜7 μm as the nonlinear medium and a 2 μm femtosecond fiber laser as the pump source, 4.5 W supercontinuum (SC) generation from 1017 to 3438 nm is obtained for a launched pump power of ˜10.48 W. The corresponding optical-to-optical conversion efficiency is about 42.9%. In addition, no any damage of the fluorotellurite fiber is observed during the operation of the above SC light source. Our results show that all-solid fluorotellurite fibers are promising nonlinear media for constructing high power mid-infrared SC light sources.

Geotechnical Diver Tools Operation and Maintenance Manual

DTIC Science & Technology

1988-10-01

taking a core with the least amount of 4. Screw piston onto bottom of piston rod; disturbance. The piston is unscrewed from lubricate U-packing seal on...corer head onto piston road. 3. Screw piston onto piston rods check U-pecking seal for proper direction (see manual). 4. Slide aore tube over piston and

General correlation for prediction of critical heat flux ratio in water cooled channels

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

Pernica, R.; Cizek, J.

1995-09-01

The paper present the general empirical Critical Heat Flux Ration (CHFR) correlation which is valid for vertical water upflow through tubes, internally heated concentric annuli and rod bundles geometries with both wide and very tight square and triangular rods lattices. The proposed general PG correlation directly predicts the CHFR, it comprises axial and radial non-uniform heating, and is valid in a wider range of thermal hydraulic conditions than previously published critical heat flux correlations. The PG correlation has been developed using the critical heat flux Czech data bank which includes more than 9500 experimental data on tubes, 7600 data onmore » rod bundles and 713 data on internally heated concentric annuli. Accuracy of the CHFR prediction, statistically assessed by the constant dryout conditions approach, is characterized by the mean value nearing 1.00 and the standard deviation less than 0.06. Moverover, a subchannel form of the PG correlations is statistically verified on Westinghouse and Combustion Engineering rod bundle data bases, i.e. more than 7000 experimental CHF points of Columbia University data bank were used.« less

Apparatus and methods for aligning holes through wheels and spacers and stacking the wheels and spacers to form a turbine rotor

DOEpatents

Berry, Robert Randolph; Palmer, Gene David; Wilson, Ian David

2000-01-01

A gas turbine rotor stacking fixture includes upstanding bolts for reception in aligned bolt holes in superposed aft disk, wheels and spacers and upstanding alignment rods received in openings of the disk, wheels and spacers during the rotor stacking assembly. The axially registering openings enable insertion of thin-walled tubes circumferentially about the rim of the rotor, with tight tolerances to the openings to provide supply and return steam for cooling buckets. The alignment rods have radial dimensions substantially less than their dimensions in a circumferential direction to allow for radial opening misalignment due to thermal expansion, tolerance stack-up and wheel-to-spacer mismatch due to rabbet mechanical growth. The circumferential dimension of the alignment rods affords tightly toleranced alignment of the openings through which the cooling tubes are installed.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

NASA Astrophysics Data System (ADS)

Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

2016-11-01

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

DOE PAGES

Williamson, R. L.; Capps, N. A.; Liu, W.; ...

2016-09-27

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) ormore » plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used in this paper to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. Finally, in comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.« less

Nuclear fuel element nut retainer cup. [PWR

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

Walton, L.A.

1977-07-19

A typical embodiment has an end fitting for a nuclear reactor fuel element that is joined to the control rod guide tubes by means of a nut plate assembly. The nut plate assembly has an array of nuts, each engaging the respective threaded end of the control rod guide tubes. The nuts, moreover, are retained on the plate during handling and before fuel element assembly by means of hollow cylindrical locking cups that are brazed to the plate and loosely circumscribe the individual enclosed nuts. After the nuts are threaded onto the respective guide tube ends, the locking cups aremore » partially deformed to prevent one or more of the nuts from working loose during reactor operation. The locking cups also prevent loose or broken end fitting parts from becoming entrained in the reactor coolant.« less

Texture evolution in Oxide Dispersion Strengthened (ODS) steel tubes during pilgering process

NASA Astrophysics Data System (ADS)

Vakhitova, E.; Sornin, D.; Barcelo, F.; François, M.

2017-10-01

Oxide Dispersion Strengthened (ODS) steels are foreseen as fuel cladding material in the coming generation of Sodium Fast Reactors (SFR). Cladding tubes are manufactured by hot extrusion and subsequent cold forming steps. In this study, a 9 wt% Cr ODS steel exhibiting α-γ phase transformation at high temperature is cold formed under industrial conditions with a large section reduction in two pilgering steps. The influence of pilgering process parameters and intermediate heat treatment on the microstructure evolution is studied experimentally using Electron Backscattering Diffraction (EBSD) and X-ray Diffraction (XRD) methods. Pilgered samples show elongated grains and a high texture formation with a preferential orientation along the rolling direction. During the heat treatment, grain morphology is recovered from elongated grains to almost equiaxed ones, while the well-known α-fiber texture presents an unexpected increase in intensity. The remarkable temperature stability of this fiber is attributed to a crystallographic structure memory effect during phase transformations.

Samus Counter Lifting Fixture

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

Stredde, H.; /Fermilab

1998-05-27

A lifting fixture has been designed to handle the Samus counters. These counters are being removed from the D-zero area and will be transported off site for further use at another facility. This fixture is designed specifically for this particular application and will be transferred along with the counters. The future use of these counters may entail installation at a facility without access to a crane and therefore a lift fixture suitable for both crane and/or fork lift usage has been created The counters weigh approximately 3000 lbs. and have threaded rods extended through the counter at the top comersmore » for lifting. When these counters were first handled/installed these rods were used in conjunction with appropriate slings and handled by crane. The rods are secured with nuts tightened against the face of the counter. The rod thread is M16 x 2({approx}.625-inch dia.) and extends 2-inch (on average) from the face of the counter. It is this cantilevered rod that the lift fixture engages with 'C' style plates at the four top comers. The strongback portion of the lift fixture is a steel rectangular tube 8-inch (vertical) x 4-inch x .25-inch wall, 130-inch long. 1.5-inch square bars are welded perpendicular to the long axis of the rectangular tube at the appropriate lift points and the 'C' plates are fastened to these bars with 3/4-10 high strength bolts -grade 8. Two short channel sections are positioned-welded-to the bottom of the rectangular tube on 40 feet centers, which are used as locators for fork lift tines. On the top are lifting eyes for sling/crane usage and are rated at 3500 lbs. safe working load each - vertical lift only.« less

Estimation of ring tensile properties of steam oxidized Zircaloy-4 fuel cladding under simulated LOCA condition

NASA Astrophysics Data System (ADS)

Shriwastaw, R. S.; Sawarn, Tapan K.; Banerjee, Suparna; Rath, B. N.; Dubey, J. S.; Kumar, Sunil; Singh, J. L.; Bhasin, Vivek

2017-09-01

The present study involves the estimation of ring tensile properties of Indian Pressurised Heavy Water Reactor (IPHWR) fuel cladding made of Zircaloy-4, subjected to experiments under a simulated loss-of-coolant-accident (LOCA) condition. Isothermal steam oxidation experiments were conducted on clad tube specimens at temperatures ranging from 900 to 1200 °C at an interval of 50 °C for different soaking periods with subsequent quenching in water at ambient temperature. The specimens, which survived quenching, were then subjected to ambient temperature ring tension test (RTT). The microstructure was correlated with the mechanical properties. The yield strength (YS) and ultimate tensile strength (UTS) increased initially with rise in oxidation temperature and time duration but then decreased with further increase in oxidation. Ductility is adversely affected with rising oxidation temperature and longer holding time. A higher fraction of load bearing phase and lower oxygen content in it ensures higher residual ductility. Cladding shows almost zero ductility behavior in RIT when load bearing phase fraction is less than 0.72 and its average oxygen concentration is greater than 0.58 wt%.

Wide angle near-field optical probes by reverse tube etching.

PubMed

Patanè, S; Cefalì, E; Arena, A; Gucciardi, P G; Allegrini, M

2006-04-01

We present a simple modification of the tube etching process for the fabrication of fiber probes for near-field optical microscopy. It increases the taper angle of the probe by a factor of two. The novelty is that the fiber is immersed in hydrofluoric acid and chemically etched in an upside-down geometry. The tip formation occurs inside the micrometer tube cavity formed by the polymeric jacket. By applying this approach, called reverse tube etching, to multimode fibers with 200/250 microm core/cladding diameter, we have fabricated tapered regions featuring high surface smoothness and average cone angles of approximately 30 degrees . A simple model based on the crucial role of the gravity in removing the etching products, explains the tip formation process.

EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

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

Saltzstein, Sylvia J.; Sorenson, Ken B.; Hanson, B. D.

The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and themore » mechanical properties of the rods will be tested and analyzed.« less

EPRI-NASA Cooperative Project on Stress Corrosion Cracking of Zircaloys. [nuclear fuel failures

NASA Technical Reports Server (NTRS)

Cubicciotti, D.; Jones, R. L.

1978-01-01

Examinations of the inside surface of irradiated fuel cladding from two reactors show the Zircaloy cladding is exposed to a number of aggressive substances, among them iodine, cadmium, and iron-contaminated cesium. Iodine-induced stress corrosion cracking (SCC) of well characterized samples of Zircaloy sheet and tubing was studied. Results indicate that a threshold stress must be exceeded for iodine SCC to occur. The existence of a threshold stress indicates that crack formation probably is the key step in iodine SCC. Investigation of the crack formation process showed that the cracks responsible for SCC failure nucleated at locations in the metal surface that contained higher than average concentrations of alloying elements and impurities. A four-stage model of iodine SCC is proposed based on the experimental results and the relevance of the observations to pellet cladding interaction failures is discussed.

M3FT-16OR0203052-Test Design for FeCrAl Alloy Tube Irradiation in HFIR

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

Terrani, Kurt A.; Petrie, Christian M.

2016-05-01

This calculation summarizes thermal analyses of a flexible rabbit design for irradiating a variety of pressurized water reactor (PWR) cladding materials (stainless steel, iron-chromium aluminum [FeCrAl], Zircaloy, and Inconel) with variable dimensions at a temperature of 350 °C in the flux trap of the High Flux Isotope Reactor (HFIR). The design can accommodate standard cladding for outer diameters (ODs) of approximately 9.50 mm with thickness ranging from 0.30 mm to 0.70 mm. The length is generally between 10 and 50 mm. The specimens contain moly inserts with a variable OD that provides the heat flux necessary to achieve the designmore » temperature with such a small fixed gas gap. The primary outer containment is an Al-6061 housing with a slightly enlarged inner diameter (ID) of 9.60 mm. The specimen temperature is controlled by determining a helium/argon gas mixture specific to the as-built specimen and housing. Variables that affect the required gas mixture are the cladding material (thermal expansion, density, heat generation rate), cladding OD, housing ID, and cladding ID. This calculation documents the analyses performed to determine required gas mixtures for a variety of scenarios.« less

Validation and evaluation of common large-area display set (CLADS) performance specification

NASA Astrophysics Data System (ADS)

Hermann, David J.; Gorenflo, Ronald L.

1998-09-01

Battelle is under contract with Warner Robins Air Logistics Center to design a Common Large Area Display Set (CLADS) for use in multiple Command, Control, Communications, Computers, and Intelligence (C4I) applications that currently use 19- inch Cathode Ray Tubes (CRTs). Battelle engineers have built and fully tested pre-production prototypes of the CLADS design for AWACS, and are completing pre-production prototype displays for three other platforms simultaneously. With the CLADS design, any display technology that can be packaged to meet the form, fit, and function requirements defined by the Common Large Area Display Head Assembly (CLADHA) performance specification is a candidate for CLADS applications. This technology independent feature reduced the risk of CLADS development, permits life long technology insertion upgrades without unnecessary redesign, and addresses many of the obsolescence problems associated with COTS technology-based acquisition. Performance and environmental testing were performed on the AWACS CLADS and continues on other platforms as a part of the performance specification validation process. A simulator assessment and flight assessment were successfully completed for the AWACS CLADS, and lessons learned from these assessments are being incorporated into the performance specifications. Draft CLADS specifications were released to potential display integrators and manufacturers for review in 1997, and the final version of the performance specifications are scheduled to be released to display integrators and manufacturers in May, 1998. Initial USAF applications include replacements for the E-3 AWACS color monitor assembly, E-8 Joint STARS graphics display unit, and ABCCC airborne color display. Initial U.S. Navy applications include the E-2C ACIS display. For these applications, reliability and maintainability are key objectives. The common design will reduce the cost of operation and maintenance by an estimated 3.3M per year on E-3 AWACS alone. It is realistic to anticipate savings of over 30M per year as CLADS is implemented widely across DoD applications. As commonality and open systems interfaces begin to surface in DoD applications, the CLADS architecture can easily and cost effectively absorb the changes, and avoid COTS obsolescence issues.

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.

Viability of thin wall tube forming of ATF FeCrAl

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

Maloy, Stuart Andrew; Aydogan, Eda; Anderoglu, Osman

Fabrication of thin walled tubing of FeCrAl alloys is critical to its success as a candidate enhanced accident-tolerant fuel cladding material. Alloys that are being investigated are Generation I and Generation II FeCrAl alloys produced at ORNL and an ODS FeCrAl alloy, MA-956 produced by Special Metals. Gen I and Gen II FeCrAl alloys were provided by ORNL and MA-956 was provided by LANL (initially produced by Special Metals). Three tube development efforts were undertaken. ORNL led the FeCrAl Gen I and Gen II alloy development and tube processing studies through drawing tubes at Rhenium Corporation. LANL received alloys frommore » ORNL and led tube processing studies through drawing tubes at Century Tubing. PNNL led the development of tube processing studies on MA-956 through pilger processing working with Sandvik Corporation. A summary of the recent progress on tube development is provided in the following report and a separate ORNL report: ORNL/TM-2015/478, “Development and Quality Assessments of Commercial Heat Production of ATF FeCrAl Tubes”.« less

Deliquification (SIC) of gas wells. Liberal District-Amoco Production Company

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

Smalley, R. Jr.

Various solutions are presented to the problem of deliquefying gas wells to achieve maximum ultimate recovery and avoid premature abandonment. Advantages and disadvantages of each method of deliquefication are discussed. The methods described include blowing up the casing, siphon strings (gas and liquids up tubing, or gas up casing-tubing annulus, and liquids up tubing), gas cycling, compression, bottomhole separators, plunger lift, and sucker rod pumping.

Retractor Tool for Brain Surgery

NASA Technical Reports Server (NTRS)

Helms, R.; Hayes, T.

1982-01-01

Proposed brain-surgery tool has an octogonal fixture for positioning latex tube over incision. Eight stainless-steel wires embedded in latex extend to hold positioning fixture. Another eight are also embedded in the latex. Concentric sleeves are successively inserted into expandable latex tube. The first sleeve is placed over a solid rod. Last sleeve is a stainless-steel tube 1 inch in diameter. It is overcoated with Teflon (or equivalent) material.

Tailoring Nd3+ emission spectrum by a neodymium-doped tellurite all-solid photonic bandgap fiber

NASA Astrophysics Data System (ADS)

Tong, Hoang Tuan; Demichi, Daisuke; Suzuki, Takenobu; Ohishi, Yasutake

2018-02-01

A tellurite all-solid photonic bandgap fiber (ASPBF) whose cladding consists of 60 high-index rods arranged periodically around a central core was successfully fabricated. The diameter of high-index rod was about 5.0 μm and the distance between the center of two adjacent high-index rods was approximately 8.0 μm. The high-index rod was made of the TeO2-Li2O-WO3-MoO3-Nb2O5 (TLWMN) glass, the cladding was made of the TeO2-ZnO-Na2O-La2O3 (TZNL) glass as the background glass material and the central core was made of TZNL glass doped with 0.5 wt% of Nd2O3. A supercontinuum light from 0.6 to 2.4 μm was coupled into the core of fiber which is 2.2 cm long to measure its transmission spectrum. High transmission bands were obtained in the vicinity of 0.75 and 1.3 μm but the transmission was suppressed in the wavelength range from 1.0 to 1.06 μm. When a titanium∶Sapphire laser source at 0.75 μm was used, the emission spectrum was obtained with two peaks at 1.06 and 1.33 μm which are attributed to the 4F3/2->4I11/2 and 4F3/2->4I13/2 transitions of Nd3+ ion, respectively. The intensities of those emission peaks were compared with those obtained from a bulk glass having the same doping concentration of Nd3+. The results showed that by using tellurite ASPBF, the intensity of the 1.06-μm emission was suppressed by one-twelfth but the intensity of the 1.33-μm emission was maintained. This feature is very advantageous to filter out the 1.06-μm emission of Nd3+ ion in order to realize practical amplifier devices at 1.3 μm.

Fabrication and characterization of an all-solid tellurite-phosphate photonic bandgap fiber.

PubMed

Cheng, Tonglei; Sakai, Yukiko; Suzuki, Takenobu; Ohishi, Yasutake

2015-05-01

We present an all-solid tellurite-phosphate photonic bandgap fiber (PBGF) with two layers of high-index rods (TeO2-Li2O-WO3-MoO3-Nb2O5, TLWMN) in the cladding (TeO2-ZnO-Li2O-K2O-Al2O3-P2O5, TZLKAP). TLWMN and TZLKAP glasses have good compatibility for fabricating the all-solid PBGF. Photonic bandgap (PBG) properties are calculated by the plane wave expansion method (PWM), and the results agree well with the measured transmission spectrum. Furthermore, the modal field patterns are measured at ∼1300 and 1520 nm, respectively. The light is confined to the core at ∼1300  nm and lost in the cladding at ∼1520  nm, which match well with the calculated modal field intensities.

Characterization of Hydrogen Embrittled Zircaloy-4 by Using a Van de Graaff Particle Accelerator

NASA Astrophysics Data System (ADS)

Budd, John

2013-04-01

On-site, dry cask storage was originally by the intended to be a short-term solution for holding spent nuclear fuel. Due to the lack of a permanent storage facility, the nuclear power industry seeks to assess the effective lifetime of the casks. One issue which could compromise cask integrity is Hydrogen embrittlement. This phenomenon occurs in the Zircaloy-4 fuel-rod cladding and is caused by the formation of Zirconium hydrides. Over time, thermal stresses caused by the heat from reactions of the stored nuclear fuel could result in significant breaches of the cladding. Our group at Texas A&M University- Kingsville is conducting experiments to aid in determining when such breaches will occur. We will irradiate samples of the alloy with protons of energies up to 400 keV using a Van de Graaff particle accelerator. Once irradiated, their properties will be characterized using scanning electron microscopy and Vickers hardness tests.

Full-scale hot cell test of an acoustic sensor dedicated to measurement of the internal gas pressure and composition of a LWR nuclear fuel rod

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

Ferrandis, J. Y.; Rosenkrantz, E.; Leveque, G.

2011-07-01

A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was carried on successfully between 2008 and 2010 on irradiated fuel rods in the LECA-STAR facility at Cadarache Centre. The acoustic sensor has been specially designed in order to provide a nondestructive technique to easily carry out the measurement of the internal gas pressure and gas composition of a LWR nuclear fuel rod. This sensor has been achieved in 2007 and is now covered by an international patent. The first positive result, concerning the device behaviour, is that the sensor-operating characteristics have notmore » been altered by a two-year exposure in the hot cell ambient. We performed the gas characterisation contained in irradiated fuel rods. The acoustic method accuracy is now {+-}5 bars on the pressure measurement result and {+-}0.3% on the evaluated gas composition. The results of the acoustic method were compared to puncture results. Another significant conclusion is that the efficiency of the acoustic method is not altered by the irradiation time, and possible modification of the cladding properties. These results make it possible to demonstrate the feasibility of the technique on irradiated fuel rods. The transducer and the associated methodology are now operational. (authors)« less

Molten core retention assembly

DOEpatents

Lampe, Robert F.

1976-06-22

Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical, imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods.

Gamma-Radiation-Induced Degradation of Actively Pumped Single-Mode Ytterbium-Doped Optical Laser - Postprint

DTIC Science & Technology

2015-01-01

evaluated using the cobalt (Co)-60 gamma irradiation facility at The Ohio State University. A radiation dose rate of 43 krad(Si)/hr was used to expose the...Table 1. Description of the optical fibers used for in-situ analysis of the radiation damage Optical fiber Core Dopant Core/cladding diameters (μm...University is a pool-type gamma irradiation facility using a common cobalt cylindrical rod irradiator submerged 20 feet into a water tank. A

Waterside corrosion of Zircaloy-clad fuel rods in a PWR environment

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

Garzarolli, F.; Jorde, D.; Manzel, R.

A data base of Zircaloy corrosion behavior under PWR operating conditions has been established from previously published reports as well as from new Kraftwerk Union (KWU) fuel examinations. The data show that the reactor environment increases the corrosion. ZrO/sub 2/ film thermal conductivity is another major factor that influences corrosion behavior. It was inferred from KWU film thickness data that the oxide film thermal conductivity may decrease once circumferential cracks develop in the layer. 57 refs.

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

DOEpatents

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

1962-08-14

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

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE PAGES

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin; ...

2017-12-22

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

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

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Two techniques enable sampling of filtered and unfiltered molten metals

NASA Technical Reports Server (NTRS)

Burris, L., Jr.; Pierce, R. D.; Tobias, K. R.; Winsch, I. O.

1967-01-01

Filtered samples of molten metals are obtained by filtering through a plug of porous material fitted in the end of a sample tube, and unfiltered samples are obtained by using a capillary-tube extension rod with a perforated bucket. With these methods there are no sampling errors or loss of liquid.

Meniscal shear stress for punching.

PubMed

Tuijthof, Gabrielle J M; Meulman, Hubert N; Herder, Just L; van Dijk, C Niek

2009-01-01

Experimental determination of the shear stress for punching meniscal tissue. Meniscectomy (surgical treatment of a lesion of one of the menisci) is the most frequently performed arthroscopic procedure. The performance of a meniscectomy is not optimal with the currently available instruments. To design new instruments, the punching force of meniscal tissue is an important parameter. Quantitative data are unavailable. The meniscal punching process was simulated by pushing a rod through meniscal tissue at constant speed. Three punching rods were tested: a solid rod of Oslash; 3.00 mm, and two hollow tubes (Oslash; 3.00-2.60 mm) with sharpened cutting edges of 0.15 mm and 0.125 mm thick, respectively. Nineteen menisci acquired from 10 human cadaveric knee joints were punched (30 tests). The force and displacement were recorded from which the maximum shear stress was determined (average added with three times the standard deviation). The maximum shear stress for the solid rod was determined at 10.2 N/mm2. This rod required a significantly lower punch force in comparison with the hollow tube having a 0.15 mm cutting edge (plt;0.01). The maximum shear stress for punching can be applied to design instruments, and virtual reality training environments. This type of experiment is suitable to form a database with material properties of human tissue similar to databases for the manufacturing industry.

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale U 3Si 2 Modeling

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

Gamble, K. A.; Hales, J. D.; Miao, Y.

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced accident tolerance when compared to traditional \\uo~fuel zircaloy clad fuel rods. One of the potential replacement fuels is uranium silicide (\\usi) for its higher thermal conductivity and uranium density. The lower melting temperature is of potential concern during postulated accident conditions. Another disadvantage for \\usi~ is the lack of experimental data under power reactor conditions. Due to the aggressive development schedule for inserting some of the potential materials into leadmore » test assemblies or rods by 2022~\\cite{bragg-sitton_2014} multiscale multiphysics modeling approaches have been used to provide insight into these materials. \\\\ \

Simulation and Measurement of Through-the-Earth, Extremely Low-Frequency Signals Using Copper-Clad Steel Ground Rods.

PubMed

Damiano, Nicholas William; Yan, Lincan; Whisner, Bruce; Zhou, Chenming

2017-01-01

The underground mining environment can greatly affect radio signal propagation. Understanding how the earth affects signal propagation is a key to evaluating communications systems used during a mine emergency. One type of communication system is through-the-earth, which can utilize extremely low frequencies (ELF). This paper presents the simulation and measurement results of recent National Institute for Occupational Safety and Health (NIOSH) research aimed at investigating current injection at ELF, and in particular, ground contact impedance. Measurements were taken at an outside surface testing location. The results obtained from modeling and measurement are characterized by electrode impedance, and the voltage received between two distant electrodes. This paper concludes with a discussion of design considerations found to affect low-frequency communication systems utilizing ground rods to inject a current into the earth.

Simulation and Measurement of Through-the-Earth, Extremely Low-Frequency Signals Using Copper-Clad Steel Ground Rods

PubMed Central

Damiano, Nicholas William; Yan, Lincan; Whisner, Bruce; Zhou, Chenming

2017-01-01

The underground mining environment can greatly affect radio signal propagation. Understanding how the earth affects signal propagation is a key to evaluating communications systems used during a mine emergency. One type of communication system is through-the-earth, which can utilize extremely low frequencies (ELF). This paper presents the simulation and measurement results of recent National Institute for Occupational Safety and Health (NIOSH) research aimed at investigating current injection at ELF, and in particular, ground contact impedance. Measurements were taken at an outside surface testing location. The results obtained from modeling and measurement are characterized by electrode impedance, and the voltage received between two distant electrodes. This paper concludes with a discussion of design considerations found to affect low-frequency communication systems utilizing ground rods to inject a current into the earth. PMID:29176916

Experimental measurement of coil-rod-coil block copolymer tracer diffusion through entangled coil homopolymers

PubMed Central

Wang, Muzhou; Timachova, Ksenia; Olsen, Bradley D.

2014-01-01

The diffusion of coil-rod-coil triblock copolymers in entangled coil homopolymers is experimentally measured and demonstrated to be significantly slower than rod or coil homopolymers of the same molecular weight. A model coil-rod-coil triblock was prepared by expressing rodlike alanine-rich α-helical polypeptides in E. coli and conjugating coillike poly(ethylene oxide) (PEO) to both ends to form coil-rod-coil triblock copolymers. Tracer diffusion through entangled PEO homopolymer melts was measured using forced Rayleigh scattering at various rod lengths, coil molecular weights, and coil homopolymer concentrations. For rod lengths, L, that are close to the entanglementh length, a, the ratio between triblock diffusivity and coil homopolymer diffusivity decreases monotonically and is only a function of L/a, in quantitative agreement with previous simulation results. For large rod lengths, diffusion follows an arm retraction scaling, which is also consistent with previous theoretical predictions. These experimental results support the key predictions of theory and simulation, suggesting that the mismatch in curvature between rod and coil entanglement tubes leads to the observed diffusional slowing. PMID:25484454

Sediment Core Extrusion Method at Millimeter Resolution Using a Calibrated, Threaded-rod.

PubMed

Schwing, Patrick T; Romero, Isabel C; Larson, Rebekka A; O'Malley, Bryan J; Fridrik, Erika E; Goddard, Ethan A; Brooks, Gregg R; Hastings, David W; Rosenheim, Brad E; Hollander, David J; Grant, Guy; Mulhollan, Jim

2016-08-17

Aquatic sediment core subsampling is commonly performed at cm or half-cm resolution. Depending on the sedimentation rate and depositional environment, this resolution provides records at the annual to decadal scale, at best. An extrusion method, using a calibrated, threaded-rod is presented here, which allows for millimeter-scale subsampling of aquatic sediment cores of varying diameters. Millimeter scale subsampling allows for sub-annual to monthly analysis of the sedimentary record, an order of magnitude higher than typical sampling schemes. The extruder consists of a 2 m aluminum frame and base, two core tube clamps, a threaded-rod, and a 1 m piston. The sediment core is placed above the piston and clamped to the frame. An acrylic sampling collar is affixed to the upper 5 cm of the core tube and provides a platform from which to extract sub-samples. The piston is rotated around the threaded-rod at calibrated intervals and gently pushes the sediment out the top of the core tube. The sediment is then isolated into the sampling collar and placed into an appropriate sampling vessel (e.g., jar or bag). This method also preserves the unconsolidated samples (i.e., high pore water content) at the surface, providing a consistent sampling volume. This mm scale extrusion method was applied to cores collected in the northern Gulf of Mexico following the Deepwater Horizon submarine oil release. Evidence suggests that it is necessary to sample at the mm scale to fully characterize events that occur on the monthly time-scale for continental slope sediments.

Sediment Core Extrusion Method at Millimeter Resolution Using a Calibrated, Threaded-rod

PubMed Central

Schwing, Patrick T.; Romero, Isabel C.; Larson, Rebekka A.; O'Malley, Bryan J.; Fridrik, Erika E.; Goddard, Ethan A.; Brooks, Gregg R.; Hastings, David W.; Rosenheim, Brad E.; Hollander, David J.; Grant, Guy; Mulhollan, Jim

2016-01-01

Aquatic sediment core subsampling is commonly performed at cm or half-cm resolution. Depending on the sedimentation rate and depositional environment, this resolution provides records at the annual to decadal scale, at best. An extrusion method, using a calibrated, threaded-rod is presented here, which allows for millimeter-scale subsampling of aquatic sediment cores of varying diameters. Millimeter scale subsampling allows for sub-annual to monthly analysis of the sedimentary record, an order of magnitude higher than typical sampling schemes. The extruder consists of a 2 m aluminum frame and base, two core tube clamps, a threaded-rod, and a 1 m piston. The sediment core is placed above the piston and clamped to the frame. An acrylic sampling collar is affixed to the upper 5 cm of the core tube and provides a platform from which to extract sub-samples. The piston is rotated around the threaded-rod at calibrated intervals and gently pushes the sediment out the top of the core tube. The sediment is then isolated into the sampling collar and placed into an appropriate sampling vessel (e.g., jar or bag). This method also preserves the unconsolidated samples (i.e., high pore water content) at the surface, providing a consistent sampling volume. This mm scale extrusion method was applied to cores collected in the northern Gulf of Mexico following the Deepwater Horizon submarine oil release. Evidence suggests that it is necessary to sample at the mm scale to fully characterize events that occur on the monthly time-scale for continental slope sediments. PMID:27585268

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

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

Rest, J.; Hofman, G.L.

1997-06-01

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

Modelling of the Gadolinium Fuel Test IFA-681 using the BISON Code

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

Pastore, Giovanni; Hales, Jason Dean; Novascone, Stephen Rhead

2016-05-01

In this work, application of Idaho National Laboratory’s fuel performance code BISON to modelling of fuel rods from the Halden IFA-681 gadolinium fuel test is presented. First, an overview is given of BISON models, focusing on UO2/UO2-Gd2O3 fuel and Zircaloy cladding. Then, BISON analyses of selected fuel rods from the IFA-681 test are performed. For the first time in a BISON application to integral fuel rod simulations, the analysis is informed by detailed neutronics calculations in order to accurately capture the radial power profile throughout the fuel, which is strongly affected by the complex evolution of absorber Gd isotopes. Inmore » particular, radial power profiles calculated at IFE–Halden Reactor Project with the HELIOS code are used. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project. Some slide have been added as an Appendix to present the newly developed PolyPole-1 algorithm for modeling of intra-granular fission gas release.« less

Advanced gray rod control assembly

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

Drudy, Keith J; Carlson, William R; Conner, Michael E

An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber tomore » enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.« less

Fundamental metallurgical aspects of axial splitting in zircaloy cladding

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

Chung, H. M.

Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10{sup 21} n cm{sup {minus}2} to 5.9 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest claddingmore » were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed.« less

The influence of strain rate and hydrogen on the plane-strain ductility of Zircaloy cladding

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

Link, T.M.; Motta, A.T.; Koss, D.A.

1998-03-01

The authors studied the ductility of unirradiated Zircaloy-4 cladding under loading conditions prototypical of those found in reactivity-initiated accidents (RIA), i.e.: near plane-strain deformation in the hoop direction (transverse to the cladding axis) at room temperature and 300 C and high strain rates. To conduct these studies, they developed a specimen configuration in which near plane-strain deformation is achieved in the gage section, and a testing methodology that allows one to determine both the limit strain at the onset of localized necking and the fracture strain. The experiments indicate that there is little effect of strain rate (10{sup {minus}3} tomore » 10{sup 2} s{sup {minus}1}) on the ductility of unhydrided Zircaloy tubing deformed under near plane-strain conditions at either room temperature or 300 C. Preliminary experiments on cladding containing 190 ppm hydrogen show only a small loss of fracture strain but no clear effect on limit strain. The experiments also indicate that there is a significant loss of Zircaloy ductility when surface flaws are present in the form of thickness imperfections.« less

Seasonal change: a device built to show why and how it happens

NASA Astrophysics Data System (ADS)

Marques, J. B.; Rocha, P. S.; Couto da Silva, T. C.

2003-08-01

Most of people believe that the seasonal change is caused by the varying distance between the Earth and the Sun as the Earth moves on its orbit around the Sun. If the varying solar distance were the cause of change of seasons, both the northern and southern hemispheres of the Earth would have the same seasonal change at the same time. But what happens is that the hemispheres have opposites seasons : when it is summer in the north it is winter in the south; when it is spring in the north it is autumn in the south, and vice-versa. The greatest factor responsible for this occurrence is the 23.5º tilt of Earth's rotational axis relative to the perpendicular to its orbit, and the fact that the Earth's rotational axis maintains a constant direction as the Earth orbits the Sun. This causes a 23.5º tilt of the celestial equator relative to the plane which contains the Sun. Because of this, the Sun can be seen as high as 23.5º above the celestial equator and low as 23.5º below it at any hemisphere. In an attempt to demonstrate why different seasons occur during the time the Earth takes to go around the Sun, a witty and original device was built using low price material such as : sewing machine pulley, wood pulley, woodbasis, wood rod, metal rod, metal bearing, metal axis, leash, nylon wire, double copper wire, isopor ball, pen tube, interrupter, lamp, nipple, plug, crank. This allow teachers to reproduce the device with their students. The idea of building a device with this goal is not new. However, a simple device which keeps the rotational axis of a figurative Earth constant in direction as it moves around the representative Sun was not found out in astronomy teaching literature. This device allows this to occur because the pen tube (which acts as the Earth's rotational axis) remains free to spin around the metal axis, kept settled. The pen tube is connected by a leash to the metal rod settled on the wood basis. When the wood system rod + pulley is turned round through a crank, the representative Earth's rotational axis (pen tube), is kept free to spin around the metal axis, maintaining its initial configuration. It is worthwhile to notice that the metal rod and the pen tube must have the same thickness. That is necessary because after a full revolution the wood system rod + pulley and the pen tube must return to the same position to keep the representative Earth's rotational axis with its initial configuration. The teacher may also explain to the students that the Earth's rotational axis really changes its positions in a very smooth way, in a period of nearly 26,000 years. This time is much greater than the lifetime of a human being, but this motion is noticed since ancient time. The device works as a helpful instrument to a teacher in any introductory astronomy course.

Simple Excitation of Standing Waves in Rubber Bands and Membranes

NASA Astrophysics Data System (ADS)

Cortel, Adolf

2004-04-01

Many methods to excite standing waves in strings, plates, membranes, rods, tubes, and soap bubbles have been described. Usually a loudspeaker or a vibrating reed is driven by the amplified output of an audio oscillator. A novel and simple method consists of using a tuning fork or a singing rod to excite transversal standing waves in stretched rubber membranes sprinkled with fine sand.

MreB filaments align along greatest principal membrane curvature to orient cell wall synthesis

PubMed Central

Szwedziak, Piotr; Wong, Felix; Schaefer, Kaitlin; Izoré, Thierry; Renner, Lars D; Holmes, Matthew J; Sun, Yingjie; Bisson-Filho, Alexandre W; Walker, Suzanne; Amir, Ariel; Löwe, Jan

2018-01-01

MreB is essential for rod shape in many bacteria. Membrane-associated MreB filaments move around the rod circumference, helping to insert cell wall in the radial direction to reinforce rod shape. To understand how oriented MreB motion arises, we altered the shape of Bacillus subtilis. MreB motion is isotropic in round cells, and orientation is restored when rod shape is externally imposed. Stationary filaments orient within protoplasts, and purified MreB tubulates liposomes in vitro, orienting within tubes. Together, this demonstrates MreB orients along the greatest principal membrane curvature, a conclusion supported with biophysical modeling. We observed that spherical cells regenerate into rods in a local, self-reinforcing manner: rapidly propagating rods emerge from small bulges, exhibiting oriented MreB motion. We propose that the coupling of MreB filament alignment to shape-reinforcing peptidoglycan synthesis creates a locally-acting, self-organizing mechanism allowing the rapid establishment and stable maintenance of emergent rod shape. PMID:29469806

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

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

Kalimullah

1994-03-01

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

Atmospheric pressure helium afterglow discharge detector for gas chromatography

DOEpatents

Rice, G.; D'Silva, A.P.; Fassel, V.A.

1985-04-05

An apparatus for providing a simple, low-frequency, electrodeless discharge system for atmospheric pressure afterglow generation. A single quartz tube through which a gas mixture is passed is extended beyond a concentric electrode positioned thereabout. A grounding rod is placed directly above the tube outlet to permit optical viewing of the discharge between the electrodes.

Atmospheric pressure helium afterglow discharge detector for gas chromatography

DOEpatents

Rice, Gary; D'Silva, Arthur P.; Fassel, Velmer A.

1986-05-06

An apparatus for providing a simple, low-frequency electrodeless discharge system for atmospheric pressure afterglow generation. A single quartz tube through which a gas mixture is passed is extended beyond a concentric electrode positioned thereabout. A grounding rod is placed directly above the tube outlet to permit optical viewing of the discharge between the electrodes.

Coal fired fluid bed module for a single elevation style fluid bed power plant

DOEpatents

Waryasz, Richard E.

1979-01-01

A fluidized bed for the burning of pulverized fuel having a specific waterwall arrangement that comprises a structurally reinforced framework of wall tubes. The wall tubes are reversely bent from opposite sides and then bonded together to form tie rods that extend across the bed to support the lateral walls thereof.

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

Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

NASA Astrophysics Data System (ADS)

Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

2016-12-01

This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

SUMMARY OF PROGRESS ON THE STUDY OF BETA TREATMENT OF URANIUM, NOVEMBER 1, 1959-AUGUST 31, 1960

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

Russell, R.B.

Variables affecting the texture and grain size of uranium during beta treatment are summarized. The study of the effect of time and temperature in the beta phase on the growth index (G3) and grain size of the final alpha product is tentatively believed to show that higher beta temperatures for short times (up to about seven minutes) tend to promote slightly more negative growth indices and that higher beta temperatures give rise to somewhat finer grain sizes. Results of studies of both Jominy end-quenched bars and several full-sized rods and tubes quenched by total immersion showed that large thermal gradientsmore » promoted negative growth indices and produced grains somewhat elongated in the direction of the thermal gradient. The effects of endcooling in full-sized pieces quenched by total immersion in cold water showed that the axial growth index is negative up to distances from the end of about half the wall thickness of tubes and about half the radial dimension of rods. The grain refinement penetrates to a lesser distance from the ends. In the radial direction the growth index for these same pieces is largely negative to a distance below the outer diameter of about midwall in two tubes studied. In the case of one tube which was studied more completely, the growth index became negative again as the inner diameter was approached. A water-quenched rod was found to have a negative growth index down to a distance from the surface of about midradius. (auth)« less

Supercontinuum generation covering the entire 0.4-5 µm transmission window in a tapered ultra-high numerical aperture all-solid fluorotellurite fiber

NASA Astrophysics Data System (ADS)

Jia, Z. X.; Yao, C. F.; Jia, S. J.; Wang, F.; Wang, S. B.; Zhao, Z. P.; Liao, M. S.; Qin, G. S.; Hu, L. L.; Ohishi, Y.; Qin, W. P.

2018-02-01

Enormous efforts have been made to realize supercontinuum (SC) generation covering the entire transmission window of fiber materials for their wide applications in many fields. Here we demonstrate ultra-broadband SC generation from 400 to 5140 nm in a tapered ultra-high numerical aperture (NA) all-solid fluorotellurite fiber pumped by a 1560 nm mode-locked fiber laser. The fluorotellurite fibers are fabricated using a rod-in-tube method. The core and cladding materials are TeO2-BaF2-Y2O3- and TeO2-modified fluoroaluminate glasses, respectively, which have large refractive index contrast and similar thermal expansion coefficients and softening temperatures. The NA at 3200 nm of the fluorotellurite fiber is about 1.11. Furthermore, tapered fluorotellurite fibers are prepared using an elongation machine. SC generation covering the entire 0.4-5 µm transmission window is achieved in a tapered fluorotellurite fiber for a pumping peak power of ~10.5 kW through synergetic control of dispersion, nonlinearity, confinement loss and other unexpected effects (e.g. the attachment of dust or water to the surface of the fiber core) of the fiber. Our results show that tapered ultra-high NA all-solid soft glass fibers have a potential for generating SC light covering their entire transmission window.

Effect of Helical Slow-Wave Circuit Variations on TWT Cold-Test Characteristics

NASA Technical Reports Server (NTRS)

Kory, Carol L.; Dayton, J. A., Jr.

1998-01-01

Recent advances in the state of the art of computer modeling offer the possibility for the first time to evaluate the effect that slow-wave structure parameter variations, such as manufacturing tolerances, have on the cold-test characteristics of helical traveling-wave tubes (TWT's). This will enable manufacturers to determine the cost effectiveness of controlling the dimensions of the component parts of the TWT, which is almost impossible to do experimentally without building a large number of tubes and controlling several parameters simultaneously. The computer code MAFIA is used in this analysis to determine the effect on dispersion and on-axis interaction impedance of several helical slow-wave circuit parameter variations, including thickness and relative dielectric constant of the support rods, tape width, and height of the metallized films deposited on the dielectric rods. Previous computer analyzes required so many approximations that accurate determinations of the effect of many relevant dimensions on tube performance were practically impossible.

Effect of Helical Slow-Wave Circuit Variations on TWT Cold-Test Characteristics

NASA Technical Reports Server (NTRS)

Kory, Carol L.; Dayton, James A., Jr.

1997-01-01

Recent advances in the state of the art of computer modeling offer the possibility for the first time to evaluate the effect that slow-wave structure parameter variations, such as manufacturing tolerances, have on the cold-test characteristics of helical traveling-wave tubes (TWT's). This will enable manufacturers to determine the cost effectiveness of controlling the dimensions of the component parts of the TWT, which is almost impossible to do experimentally without building a large number of tubes and controlling several parameters simultaneously. The computer code MAFIA is used in this analysis to determine the effect on dispersion and on-axis interaction impedance of several helical slow-wave circuit parameter variations, including thickness and relative dielectric constant of the support rods, tape width, and height of the metallized films deposited on the dielectric rods. Previous computer analyses required so many approximations that accurate determinations of the effect of many relevant dimensions on tube performance were practically impossible.

Effect of Helical Slow-Wave Circuit Variations on TWT Cold-Test Characteristics

NASA Technical Reports Server (NTRS)

Kory, Carol L.; Dayton, James A., Jr.

1998-01-01

Recent advances in the state of the art of computer modeling offer the possibility for the first time to evaluate the effect that slow-wave structure parameter variations, such'as manufacturing tolerances, have on the cold-test characteristics of helical traveling-wave tubes (TWT's). This will enable manufacturers to determine the cost effectiveness of controlling the dimensions of the component parts of the TWT, which is almost impossible to do experimentally without building a large number of tubes and controlling several parameters simultaneously. The computer code MAxwell's equations by the Finite Integration Algorithm (MAFIA) is used in this analysis to determine the effect on dispersion and on-axis interaction impedance of several helical slow-wave circuit parameter variations, including thickness and relative dielectric constant of the support rods, tape width, and height of the metallized films deposited on the dielectric rods. Previous computer analyzes required so many approximations that accurate determinations of the effect of many relevant dimensions on tube performance were practically impossible.

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

Huang, Chuanfu; Kresin, Vitaly V., E-mail: kresin@usc.edu

This note describes a system for transferring a load of high purity lithium metal into a molecular or cluster beam source. A hot loading vessel is thoroughly baked out while empty and overpressured with argon. A clean Li rod is then dropped in through a long narrow tube. The thoroughly degassed interior of the vessel and the rapid melting of the inserted rod facilitate contamination-free transfer of the highly reactive liquid metal into the source oven.

Design and analysis of the radiator structure for space power systems

NASA Technical Reports Server (NTRS)

Dauterman, W. H.; Montgomery, L. D.

1973-01-01

The design, analysis, fabrication, and development of the 5-kWe radiator structure are shown. Thermal performance, meteoroid protection, structural capability during launch, development testing and space operation, material evaluation, and the configuration selection are described. The fin-tube development program depends on the relative values of the thermal coefficients of expansion. The initial selection of aluminum fins and Type 316 stainless-steel tubes was based on previous experience; however, the large differential in their expansion rates showed that an alternate, more compatible, combination was needed. Copper, stainless-steel-clad copper, boron-impregnated aluminum, and an independent radiator with a titanium structure were all considered as alternate materials. The final selection was Lockalloy fins with Type 304 stainless-steel D tubes.

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.

Manifold tool guide

DOEpatents

Djordjevic, A.

1982-07-08

A tool guide that makes possible the insertion of cleaning and/or inspection tools into a manifold pipe that will dislocate and extract the accumulated sediment in such manifold pipes. The tool guide basically comprises a right angled tube (or other angled tube as required) which can be inserted in a large tube and locked into a radially extending cross pipe by adjustable spacer rods and a spring-loaded cone, whereby appropriate cleaning tools can be inserted into to cross pipe for cleaning, inspection, etc.

NEUTRONIC REACTOR CONSTRUCTION

DOEpatents

Vernon, H.C.; Goett, J.J.

1958-09-01

A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

Manifold tool guide

DOEpatents

Djordjevic, A.

1983-12-27

A tool guide is described that makes possible the insertion of cleaning and/or inspection tools into a manifold pipe that will dislocate and extract the accumulated sediment in such manifold pipes. The tool guide basically comprises a right angled tube (or other angled tube as required) which can be inserted in a large tube and locked into a radially extending cross pipe by adjustable spacer rods and a spring-loaded cone, whereby appropriate cleaning tools can be inserted into the cross pipe for cleaning, inspection, etc. 3 figs.

Manifold tool guide

DOEpatents

Djordjevic, Aleksandar

1983-12-27

A tool guide that makes possible the insertion of cleaning and/or inspection tools into a manifold pipe that will dislocate and extract the accumulated sediment in such manifold pipes. The tool guide basically comprises a right angled tube (or other angled tube as required) which can be inserted in a large tube and locked into a radially extending cross pipe by adjustable spacer rods and a spring-loaded cone, whereby appropriate cleaning tools can be inserted into to cross pipe for cleaning, inspection, etc.

The Loss Cone Imager (LCI)

DTIC Science & Technology

2006-07-24

at this time that the PMO provided the current LCI configuration (integrated DPU/HST on the PM, the RSH units on a composite tube , later to become...approximated by parallel planes of 1000 A of aluminum on a 0.5 mil mylar foil in front of a 1000 tim (or 1500 tim) silicon detector, detector A...above manufacturers form these alloys in tube form. In order to produce a tube , a rod would be formed and then a hold bored through the center. Some of

Site Investigations with the Site Characterization and Analysis Penetrator System at Fort Dix, New Jersey

DTIC Science & Technology

1993-07-01

rod system or through a tremie tube ; both procedures were used interchangeably at Fort Dix to demon- strate the efficiency and effectiveness of each...allows delivery through either a l/4-in.-diam grout tube or a 3/8-in.-diam rout tube . The grout used at Fort Dix consisted of a mixture of water and... microfine , blended Portland cement (Lehigh Geocem’, Leeds, Alabama). The grout is a suspension of a uniformly produced cement clinker interground with

CUTTING AND WEDGING JACKET REMOVER

DOEpatents

Freedman, M.; Raynor, S.

1959-04-01

A tool is presented for stripping cladded jackets from fissionable fuel elements. The tool is a tube which fits closely around the jacket and which has two cutting edges at opposite sides of one end. These cutting edges are adjusted to penetrate only the jacket so that by moving the edges downward the jacket is cut into two pieces.

Modeling Explosive Cladding of Metallic Liners to Gun Tubes

DTIC Science & Technology

2010-01-01

a Jones Wilkins Lee ( JWL ) equation of state was parameterized using nonlinear optimization (ref. 8) and scaling the empirical v2£ for other volume...expansions based on TNT. The JWL equation of state is ( ,. A i -RiV* , GJE RiV* V IS./K v f +7~* (2) where P is pressure, V is

Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

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

Ferrandis, J.Y.; Leveque, G.; Rosenkrantz, E.

2015-07-01

First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes amore » main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was conducted successfully between 2008 and 2010 on 5 high burn-up MOX fuel rods and 2 very high burn-up UO{sub 2} fuel rods in LECA Facility at Cadarache Centre. An improvement of this sensor has been proposed, allowing us to divide by two the uncertainty on the pressure measurement. In the case of hot-cell measurements, viscous liquid can be used to couple the sensor with the rod. For gas content with a pressure exceeding 15 bars and a 10% Xe/Kr ratio, such coupling may reduce relative acoustic method accuracy by ±7% for pressure measurement result and ±0.25 % for the assessment of gas composition. These results make it possible to demonstrate the feasibility of the technique on LWR fuel rods. The transducer and the associated methodology are now operational for non-destructive measurements in hot lab facilities and allow characterising the fission gas without puncturing the fuel rods. Up to now, any other non-destructive method can be proposed. A next step will be the development of an industrial application in a fuel storage pool in order to perform a large number of measurements on a fuel assembly in nuclear plants.« less

APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND

DOEpatents

Treshow, M.

1963-07-23

A method of controlling a fuel-rod-in-tube-type boilingwater reactor having nozzles at the point of water entry into the tube is described. Water is pumped into the nozzles by an auxiliary pump operated by steam from an interstage position of the associated turbine, so that the pumping speed is responsive to turbine demand. (AEC)

78 FR 56655 - Foreign-Trade Zone (FTZ) 203-Moses Lake, Washington; Notification of Proposed Production Activity...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-13

... choose the duty rate during customs entry procedures that applies to fuel rod assemblies (duty rate--3.3... plates; channels; spacers; end caps; springs; top and bottom nozzles; cages; guide tubes; and tubes, bar stocks, wires and plates of zirconium (duty rate ranges from 3.3 to 3.7%). Public comment is invited from...

Harmonic Resonances in Metal Rods--Easy Experimentation with a Smartphone and Tablet PC

ERIC Educational Resources Information Center

Hirth, Michael; Gröber, Sebastian; Kuhn, Jochen; Müller, Andreas

2016-01-01

A variety of experiments with smartphones and tablet PCs allow a precise analysis of acoustic phenomena. For example, we recently described an experiment in which a noise signal simultaneously triggered harmonic acoustic resonances in the air column of a tube open at both ends and a tube closed at one end. The measured resonant frequencies were…

Cavitation phenomena in mechanical heart valves: studied by using a physical impinging rod system.

PubMed

Lo, Chi-Wen; Chen, Sheng-Fu; Li, Chi-Pei; Lu, Po-Chien

2010-10-01

When studying mechanical heart valve cavitation, a physical model allows direct flow field and pressure measurements that are difficult to perform with actual valves, as well as separate testing of water hammer and squeeze flow effects. Movable rods of 5 and 10 mm diameter impinged same-sized stationary rods to simulate squeeze flow. A 24 mm piston within a tube simulated water hammer. Adding a 5 mm stationary rod within the tube generated both effects simultaneously. Charged-coupled device (CCD) laser displacement sensors, strobe lighting technique, laser Doppler velocimetry (LDV), particle image velocimetry (PIV) and high fidelity piezoelectric pressure transducers measured impact velocities, cavitation images, squeeze flow velocities, vortices, and pressure changes at impact, respectively. The movable rods created cavitation at critical impact velocities of 1.6 and 1.2 m/s; squeeze flow velocities were 2.8 and 4.64 m/s. The isolated water hammer created cavitation at 1.3 m/s piston speed. The combined piston and stationary rod created cavitation at an impact speed of 0.9 m/s and squeeze flow of 3.2 m/s. These results show squeeze flow alone caused cavitation, notably at lower impact velocity as contact area increased. Water hammer alone also caused cavitation with faster displacement. Both effects together were additive. The pressure change at the vortex center was only 150 mmHg, which cannot generate the magnitude of pressure drop required for cavitation bubble formation. Cavitation occurred at 3-5 m/s squeeze flow, significantly different from the 14 m/s derived by Bernoulli's equation; the temporal acceleration of unsteady flow requires further study.

Fuel subassembly leak test chamber for a nuclear reactor

DOEpatents

Divona, Charles J.

1978-04-04

A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

Extending Mode Areas of Single-mode All-solid Photonic Bandgap Fibers

DTIC Science & Technology

2015-04-02

T. Tunnermann, R. Iliew, F. Lederer, J. Broeng, G. Vienne, A. Petersson, and C. Jakobsen, “High-power air-clad large-mode-area photonic crystal ...Yvernault, and F. Salin, “Extended single-mode photonic crystal fiber lasers,” Opt. Express 14(7), 2715–2720 (2006). 10. L. Dong, T. Wu, H. McKay, L. Fu...progress in mode area scaling of optical fibers. One notable area is in photonic crystal fibers (PCF) [3–5, 8, 9]. The short straight PCF rods used in

Sensitivity analysis of FeCrAl cladding and U3Si2 fuel under accident conditions

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

Gamble, Kyle Allan Lawrence; Hales, Jason Dean

2016-08-01

The purpose of this milestone report is to highlight the results of sensitivity analyses performed on two accident tol- erant fuel concepts: U3Si2 fuel and FeCrAl cladding. The BISON fuel performance code under development at Idaho National Laboratory was coupled to Sandia National Laboratories’ DAKOTA software to perform the sensitivity analyses. Both Loss of Coolant (LOCA) and Station blackout (SBO) scenarios were analyzed using main effects studies. The results indicate that for FeCrAl cladding the input parameters with greatest influence on the output metrics of interest (fuel centerline temperature and cladding hoop strain) during the LOCA were the isotropic swellingmore » and fuel enrichment. For U3Si2 the important inputs were found to be the intergranular diffusion coefficient, specific heat, and fuel thermal conductivity. For the SBO scenario, Young’s modulus was found to be influential in FeCrAl in addition to the isotropic swelling and fuel enrichment. Contrarily to the LOCA case, the specific heat of U3Si2 was found to have no effect during the SBO. The intergranular diffusion coefficient and fuel thermal conductivity were still found to be of importance. The results of the sensitivity analyses have identified areas where further research is required including fission gas behavior in U3Si2 and irradiation swelling in FeCrAl. Moreover, the results highlight the need to perform the sensitivity analyses on full length fuel rods for SBO scenarios.« less

Magnetic field sensing based on tilted fiber Bragg grating coated with nanoparticle magnetic fluid

NASA Astrophysics Data System (ADS)

Yang, Dexing; Du, Lei; Xu, Zengqi; Jiang, Yajun; Xu, Jian; Wang, Meirong; Bai, Yang; Wang, Haiyan

2014-02-01

A magnetic field sensor based on a tilted fiber Bragg grating (TFBG) coated with magnetic fluid is proposed and demonstrated experimentally. The sensing element is made by injecting the magnetic fluid into a capillary tube which contains a TFBG. The resonant wavelengths of the cladding modes of TFBG shift by varying the magnetic field which is perpendicular to the axis of TFBG. The results indicate that the resonant wavelength shifts of the cladding modes show a nonlinear dependence on the magnetic field. As the magnetic field increases to 32 mT, the largest resonant wavelength shift reaches to 106 pm. Moreover, this sensor shows good repeatability when it is used for magnetic field sensing.

Aspects of forming metal-clad melt-processed Y-Ba-Cu-O tapes

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

Kozlowski, G.; Oberly, C.E.; Ho, J.

1991-03-01

This paper reports on melt-processing of Y-Ba-Cu-O superconductor in a usable form for magnet winding which requires the development of a cladding with demanding properties. Numerous recent efforts in cold forming Bi-based superconductor tapes have been successful because a silver tube can be used to constrain the ceramic material, which is sintered at much lower temperature than the Y-Ba-Cu-O. Typical high temperature metals which can be used to encase Y-Ba-Cu-O during sintering do not permit ready diffusion of oxygen as silver does. Recently, the full or partial recovery of superconductivity has been achieved in transition-metal- doped Y-Ba-Cu-O due to themore » partial-melt processing.« less

Electrically isolated, high melting point, metal wire arrays and method of making same

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

Simpson, John T.; Cunningham, Joseph P.; D'Urso, Brian R.

2016-01-26

A method of making a wire array includes the step of providing a tube of a sealing material and having an interior surface, and positioning a wire in the tube, the wire having an exterior surface. The tube is heated to soften the tube, and the softened tube is drawn and collapsed by a mild vacuum to bring the interior surface of the tube into contact with the wire to create a coated wire. The coated wires are bundled. The bundled coated wires are heated under vacuum to fuse the tube material coating the wires and create a fused rodmore » with a wire array embedded therein. The fused rod is cut to form a wire array. A wire array is also disclosed.« less

Ohno continuous casting

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

Soda, H.; McLean, A.; Motoyasu, G.

1995-04-01

Production of net-shape products directly from the liquid is an attractive manufacturing route for alloys that are difficult to process or that cannot be rolled, drawn, or extruded. Developed at the Chiba Institute of Technology in Japan, the Ohno Continuous Casting (OCC) approach not only provides significant cost savings, but also has the potential to create new products. OCC process equipment includes a melting furnace, crucible, mold level-control block, cooling device, and pinch rolls. OCC is currently used to produce copper rods and wires for audio and video cables, and aluminum alloy welding rods for hard-surfacing applications. For example, Mitsuimore » Engineering and Ship Building Co. has used OCC to produce copper tubing products with internal fins and partitions for applications such as heat exchanger tubes and induction coils.« less

The fractalline properties of experimentally simulated PWR fuel crud

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Mishra, V. K.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Mota, G.; Hussey, D.; Wang, G.; Byers, W. A.; Short, M. P.

2018-02-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures.

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

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

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

Refractory clad transient internal probe for magnetic field measurements in high temperature plasmas

NASA Astrophysics Data System (ADS)

Kim, Hyundae; Cellamare, Vincent; Jarboe, Thomas R.; Mattick, Arthur T.

2005-05-01

The transient internal probe (TIP) is a diagnostic for local internal field measurements in high temperature plasmas. A verdet material, which rotates the polarization angle of the laser light under magnetic fields, is launched into a plasma at about 1.8km/s. A linearly polarized Ar+ laser illuminates the probe in transit and the light retroreflected from the probe is analyzed to determine the local magnetic field profiles. The TIP has been used for magnetic field measurements on the helicity injected torus where electron temperature Te⩽80eV. In order to apply the TIP in higher temperature plasmas, refractory clad probes have been developed utilizing a sapphire tube, rear disc, and a MgO window on the front. The high melting points of these refractory materials should allow probe operation at plasma electron temperatures up to Te˜300eV. A retroreflecting probe has also been developed using "catseye" optics. The front window is replaced with a plano-convex MgO lens, and the back surface of the probe is aluminized. This approach reduces spurious polarization effects and provides refractory cladding for the probe entrance face. In-flight measurements of a static magnetic field demonstrate the ability of the clad probes to withstand gun-launch acceleration, and provide high accuracy measurements of magnetic field.

In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

DOE PAGES

Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; ...

2015-08-25

Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filteringmore » unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.« less

"Teaching" an Industrial Robot To Spray

NASA Technical Reports Server (NTRS)

Evans, A. R.; Sweet, G. K.

1982-01-01

Teaching device, consisting of spacer rod or tube with three-pointed tip and line level, is used during pattern "teach-in" to make sure that robot manipulator holds spray gun perpendicular to surface to be sprayed and at right distance from it. For slanted surfaces angle adapter is added between spacer rod and line-level indicator. Angle is determined by slope of surface to be sprayed, thus allowing a perpendicular spray pattern against even slanted surfaces.

High-temperature, high-pressure optical cell

NASA Technical Reports Server (NTRS)

Harris, R. P. (Inventor); Holland, L. R. (Inventor); Smith, R. E. (Inventor)

1986-01-01

The invention is an optical cell for containment of chemicals under conditions of high temperature and high pressure. The cell is formed of a vitreous silica tube, two optical windows comprising a vitreous silica rod inserted into the ends of a tube, and fused into position in the tube ends. Windows are spaced apart to form a cavity enclosed by the tube and the windows. A hole is drilled radially through the tube and into the cavity. Another vitreous silica tube is fused to the silica tube around the hole to form the stem, which is perpendicular to the long axis of the tube. The open end of the stem is used to load chemicals into the cavity. Then the stem may be sealed, and if desired, it may be shortened in order to reduce the volume of the cavity, which extends into the stem.

Vertebral coplanar alignment: a standardized technique for three dimensional correction in scoliosis surgery: technical description and preliminary results in Lenke type 1 curves.

PubMed

Vallespir, Gabriel Pizà; Flores, Jesús Burgos; Trigueros, Ignacio Sanpera; Sierra, Eduardo Hevia; Fernández, Pedro Doménech; Olaverri, Juan Carlos Rodríguez; Alonso, Manuel García; Galea, Rafael Ramos; Francisco, Antonio Pérez; Rodríguez de Paz, Beatriz; Carbonell, Pedro Gutiérrez; Thomas, Javier Vicente; López, José Luís González; Paulino, José Ignacio Maruenda; Pitarque, Carlos Barrios; García, Oscar Riquelme

2008-06-15

Prospective multicentric study. To present the preliminary results of an innovative method for standardized correction of scoliosis, vertebral coplanar alignment (VCA), based on a novel concept: the relocation of vertebral axis in a single plane. Normal standing spine has no rotation in coronal or transverse planes, therefore X and Z axis of vertebrae are in the same plane: they are coplanar. VCA intends to relocate these axis in one plane, correcting rotation and translation, while X axis are returned to its normal posterior divergence in sagittal plane in thoracic spine. Twenty-five consecutive adolescent idiopathic scoliosis patients (Lenke type 1) underwent posterior surgery with segmental pedicle screw fixation. Slotted tubes were attached to convex side screws. Two longitudinal rods were inserted through the end of tubes. Then, they were separated along the slots, driving the tubes into one plane, making the axis of the vertebrae coplanar and thus correcting transverse rotation and coronal translation. To obtain kyphosis, distal ends of the tubes were spread in thoracic spine. Correction was maintained by locking a definitive rod in the concave side, then tubes were retrieved and the convex side rod, inserted and tightened. Correction was assessed on preoperative and postoperative full-spine standing radiograph. Vertebral rotation was measured on computed tomography-scan and magnetic resonance imaging. Preoperative average thoracic curves of 61 degrees were corrected to 16 degrees (73%). Preoperative average thoracolumbar curves of 39 degrees were corrected to 12 degrees (70%). Preoperative average thoracic apical rotation of 24 degrees was corrected to 11 degrees (56%). Preoperative average thoracic kyphosis of 18 degrees remained unchanged after surgery; however, no patients had kyphosis <10 degrees after surgery. Rib hump improved from 30 to 11 mm (65%). There were no perioperative complications. VCA provided excellent correction of coronal and transverse planes with normalization of thoracic kyphosis in Lenke type 1 adolescent idiopathic scoliosis surgery.

40 CFR 420.111 - Specialized definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... steel products such as coiled wire, rods, and tubes in discrete batches or bundles. (b) The term continuous means those alkaline cleaning operations which process steel products other than in discrete...

Eulerian Formulation of Spatially Constrained Elastic Rods

NASA Astrophysics Data System (ADS)

Huynen, Alexandre

Slender elastic rods are ubiquitous in nature and technology. For a vast majority of applications, the rod deflection is restricted by an external constraint and a significant part of the elastic body is in contact with a stiff constraining surface. The research work presented in this doctoral dissertation formulates a computational model for the solution of elastic rods constrained inside or around frictionless tube-like surfaces. The segmentation strategy adopted to cope with this complex class of problems consists in sequencing the global problem into, comparatively simpler, elementary problems either in continuous contact with the constraint or contact-free between their extremities. Within the conventional Lagrangian formulation of elastic rods, this approach is however associated with two major drawbacks. First, the boundary conditions specifying the locations of the rod centerline at both extremities of each elementary problem lead to the establishment of isoperimetric constraints, i.e., integral constraints on the unknown length of the rod. Second, the assessment of the unilateral contact condition requires, in principle, the comparison of two curves parametrized by distinct curvilinear coordinates, viz. the rod centerline and the constraint axis. Both conspire to burden the computations associated with the method. To streamline the solution along the elementary problems and rationalize the assessment of the unilateral contact condition, the rod governing equations are reformulated within the Eulerian framework of the constraint. The methodical exploration of both types of elementary problems leads to specific formulations of the rod governing equations that stress the profound connection between the mechanics of the rod and the geometry of the constraint surface. The proposed Eulerian reformulation, which restates the rod local equilibrium in terms of the curvilinear coordinate associated with the constraint axis, describes the rod deformed configuration by means of either its relative position with respect to the constraint axis (contact-free segments) or its angular position on the constraint surface (continuous contacts.) This formulation circumvents both drawbacks that afflict the conventional Lagrangian approach associated with the segmentation strategy. As the a priori unknown domain, viz. the rod length, is substituted for the known constraint axis, the free boundary problem and the associated isoperimetric constraints are converted into a classical two-point boundary value problem. Additionally, the description of the rod deflection by means of its eccentricity with respect to the constraint axis trivializes the assessment of the unilateral contact condition. Along continuous contacts, this formulation expresses the strain variables, measuring the rod change of shape, in terms of the geometric invariants of the constraint surface, and emphasizes the influence of the constraint local geometry on the reaction pressure. Formalizing the segmentation strategy, a computational model that exploits the Eulerian formulation of the rod governing equations is devised. To solve the quasi-static deflection of elastic rods constrained inside or around a tube-like surface, this computational model identifies the number of contacts, their nature (either discrete or continuous), and the rod configuration at the connections that satisfies the unilateral contact condition and preserves the rod integrity along the sequence of elementary problems.

Studying Hooke's Law by Using a Pogo Stick

ERIC Educational Resources Information Center

Silva, Nicolas

2011-01-01

Perhaps the pogo stick was little Robert Hooke's favorite childhood toy, consisting of a stiff spring inserted in a tube fixed at the upper end and connected to a moveable rod at the other. Hand grips and a foot rest are connected to the tube. The idea is to jump on it taking advantage of the force provided by the spring when it is compressed.…

Thermal Modelling of Various Thermal Barrier Coatings in a High Flux Rocket Engine

NASA Technical Reports Server (NTRS)

Nesbitt, James A.

1998-01-01

A thermal model was developed to predict the thermal response of coated and uncoated tubes tested in a H2/O2 rocket engine. Temperatures were predicted for traditional APS ZrO2-Y2O3 thermal barrier coatings, as well as APS and LPPS ZrO2-Y2O3/NiCrAlY cermet coatings. Good agreement was observed between predicted and measured metal temperatures at locations near the tube surface or at the inner tube wall. The thermal model was also used to quantitatively examine the effect of various coating system parameters on the temperatures in the substrate and coating. Accordingly, the effect of the presence a metallic bond coat and the effect of radiation from the surface of the ceramic layer were examined. In addition, the effect of a variation in the values of the thermal conductivity of the ceramic layer was also investigated. It was shown that a variation in the thermal conductivity of the ceramic layer, on the order of that reported in the literature for plasma sprayed ZrO2-Y2O3 coatings, can result in temperature differences in the substrate greater than 100 C, a much greater effect than that due to the presence of a bond coat or radiation from the ceramic layer. The thermal model was also used to predict the thermal response of a coated rod in order to quantify the difference in the metal temperatures between the two substrate geometries in order to explain the previously-observed increased life of coatings on rods over that on tubes. It was shown that for the short duration testing in the rocket engine, the temperature in a tube could exceed that in a rod by more than 100 C. Lastly, a two-dimensional model was developed to evaluate the effect of tangential heat transfer around the tube and its impact on reducing the stagnation point temperature. It was also shown that tangential heat transfer does not significantly reduce the stagnation point temperature, thus allowing application of a simpler, one-dimensional model for comparing measured and predicted stagnation point temperatures.

TRANSURANUS: a fuel rod analysis code ready for use

NASA Astrophysics Data System (ADS)

Lassmann, K.

1992-06-01

TRANSURANUS is a computer program for the thermal and mechanical analysis of fuel rods in nuclear reactors and was developed at the European Institute for Transuranium Elements (TUI). The TRANSURANUS code consists of a clearly defined mechanical-mathematical framework into which physical models can easily be incorporated. Besides its flexibility for different fuel rod designs the TRANSURANUS code can deal with very different situations, as given for instance in an experiment, under normal, off-normal and accident conditions. The time scale of the problems to be treated may range from milliseconds to years. The code has a comprehensive material data bank for oxide, mixed oxide, carbide and nitride fuels, Zircaloy and steel claddings and different coolants. During its development great effort was spent on obtaining an extremely flexible tool which is easy to handle, exhibiting very fast running times. The total development effort is approximately 40 man-years. In recent years the interest to use this code grew and the code is in use in several organisations, both research and private industry. The code is now available to all interested parties. The paper outlines the main features and capabilities of the TRANSURANUS code, its validation and treats also some practical aspects.

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).

Study of iodine migration in zirconia using stable and radioactive ion implantation

NASA Astrophysics Data System (ADS)

Chevarier, N.; Brossard, F.; Chevarier, A.; Crusset, D.; Moncoffre, N.

1998-03-01

The large uranium fission cross section leading to iodine and the behaviour of this element in the cladding tube during energy production and afterwards during waste storage is a crucial problem, especially for 129I which is a very long half-life isotope ( T = 1.59 × 10 7yr). Since a combined external and internal oxidation of the zircaloy cladding tube occurs during the reactor processing, iodine diffusion parameters in zirconia are needed. In order to obtain these data, stable iodine atoms were first introduced by ion implantation into zirconia with an energy of 200 keV and a dose equal to 8 × 10 15at cm -2. Diffusion profiles were measured using 3 MeV alpha-particle Rutherford Backscattering Spectrometry at each step of the annealing procedure between 700°C and 900°C. In such experiments a reduced iodine concentration was observed, which correlated to a diffusion-like process. Similar analysis has been performed using radioactive 131I implanted at a very low dose of 10 9 at cm -2. In this case the iodine release is deduced from gamma-ray spectroscopy measurements. The results are discussed in this paper.

Strength of SiCf-SiCm composite tube under uniaxial and multiaxial loading

NASA Astrophysics Data System (ADS)

Shapovalov, Kirill; Jacobsen, George M.; Alva, Luis; Truesdale, Nathaniel; Deck, Christian P.; Huang, Xinyu

2018-03-01

The authors report mechanical strength of nuclear grade silicon carbide fiber reinforced silicon carbide matrix composite (SiCf-SiCm) tubing under several different stress states. The composite tubing was fabricated via a Chemical Vapor Infiltration (CVI) process, and is being evaluated for accident tolerant nuclear fuel cladding. Several experimental techniques were applied including uniaxial tension, elastomer insert burst test, open and closed end hydraulic bladder burst test, and torsion test. These tests provided critical stress and strain values at proportional limit and at ultimate failure points. Full field strain measurements using digital image correlation (DIC) were obtained in order to acquire quantitative information on localized deformation during application of stress. Based on the test results, a failure map was constructed for the SiCf-SiCm composites.

Internal absorber solar collector

DOEpatents

Sletten, Carlyle J.; Herskovitz, Sheldon B.; Holt, F. S.; Sletten, E. J.

1981-01-01

Thin solar collecting panels are described made from arrays of small rod collectors consisting of a refracting dielectric rod lens with an absorber imbedded within it and a reflecting mirror coated on the back side of the dielectric rod. Non-tracking collector panels on vertical walls or roof tops receive approximately 90% of solar radiation within an acceptance zone 60.degree. in elevation angle by 120.degree. or more in the azimuth sectors with a collector concentration ratio of approximately 3.0. Miniaturized construction of the circular dielectric rods with internal absorbers reduces the weight per area of glass, plastic and metal used in the collector panels. No external parts or insulation are needed as heat losses are low due to partial vacuum or low conductivity gas surrounding heated portions of the collector. The miniature internal absorbers are generally made of solid copper with black selective surface and the collected solar heat is extracted at the collector ends by thermal conductivity along the absorber rods. Heat is removed from end fittings by use of liquid circulants. Several alternate constructions are provided for simplifying collector panel fabrication and for preventing the thermal expansion and contraction of the heated absorber or circulant tubes from damaging vacuum seals. In a modified version of the internal absorber collector, oil with temperature dependent viscosity is pumped through a segmented absorber which is now composed of closely spaced insulated metal tubes. In this way the circulant is automatically diverted through heated portions of the absorber giving higher collector concentration ratios than theoretically possible for an unsegmented absorber.

40 CFR 420.91 - Specialized definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... wire, rods, and tubes in discrete batches or bundles. (f) The term continuous means those pickling operations which process steel products other than in discrete batches or bundles. (g) The term acid recovery...

NUCLEAR REACTOR UNLOADING APPARATUS

DOEpatents

Leverett, M.C.; Howe, J.P.

1959-01-20

An unloading device is described for a heterogeneous reactor of the type wherein the fuel elements are in the form of cylindrical slugs and are disposed in horizontal coolant tubes which traverse the reactor core, coolant fluid being circulated through the tubes. The coolant tubes have at least two inwardly protruding ribs from their lower surfaces to support the slugs in spaced relationship to the inside walls of the tubes. The unloading device consists of a ribbon-like extractor member insertable into the coolant tubes in the space between the ribs and adapted to slide under the fuel slugs thereby raising them off of the ribs and forming a slideway for removing them from the reactor. The fuel slugs are ejected by being forced out of the tubes by incoming new fuel slugs or by a push rod insentable through the inlet end of the fuel tubes.

Effects of high thermal and high fast fluences on the mechanical properties of type 6061 aluminum in the HFBR

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

Weeks, J.R.; Czajkowski, C.J.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is an epithermal, externally moderated (by D/sub 2/O) facility designed to produce neutron beams for research. Type 6061 T-6 aluminum was used for the beam tubes, pressure vessel, fuel cladding, and most other components in the high flux area. The HFBR has operated since 1965. The epithermal, external moderation of the HFBR means that materials irradiated in different areas of the facility receive widely different flux spectra. Thus, specimens from a control rod drive follower tube (CRDF) have received 1.5 /times/ 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV)more » and 3.2 /times/ 10/sup 23/ n/cm/sup 2/ thermal fluence, while those from a vertical thimble flow shroud received 1.9 /times/ 10/sup 23/ n/cm/sup 2/ (E > 0.1 MeV) and 1.0 /times/ 10/sup 23/ n/cm/sup 2/ thermal. These numbers correspond to fast to thermal fluence ratios ranging from 0.05 to 1.9. Irradiations are occurring at approximately 333/degree/K. The data indicate that the increase in tensile strength and decrease in ductility result primarily from the thermal fluence, i.e., the transmutation of aluminum to silicon. These effects appear to be saturating at fluences above approximately 1.8 /times/ 10/sup 23/ n/cm/sup 2/ thermal at values of 90,000 psi (6700 Kg/mm/sup 2/) and 9%, respectively. The specimens receiving the highest fluence ratios appear to have less increase in tensile strength and less decrease in ductility than specimens with a lower fast to thermal fluence ratio and the same thermal fluence, suggesting a possible beneficial effect of the high energy neutrons in preventing formation of silicon crystallites. 7 refs., 11 figs., 3 tabs.« less

Process for producing fine and ultrafine filament superconductor wire

DOEpatents

Kanithi, H.C.

1992-02-18

A process for producing a superconductor wire made up of a large number of round monofilament rods is provided for, comprising assembling a multiplicity of round monofilaments inside each of a multiplicity of thin wall hexagonal tubes and then assembling a number of said thin wall hexagonal tubes within an extrusion can and subsequently consolidating, extruding and drawing the entire assembly down to the desired wire size. 8 figs.

Process for producing fine and ultrafine filament superconductor wire

DOEpatents

Kanithi, Hem C.

1992-01-01

A process for producing a superconductor wire made up of a large number of round monofilament rods is provided for, comprising assembling a multiplicity of round monofilaments inside each of a multiplicity of thin wall hexagonal tubes and then assembling a number of said thin wall hexagonal tubes within an extrusion can and subsequently consolidating, extruding and drawing the entire assembly down to the desired wire size.

Surface treatment to form a dispersed Y2O3 layer on Zircaloy-4 tubes

NASA Astrophysics Data System (ADS)

Jung, Yang-Il; Kim, Hyun-Gil; Guim, Hwan-Uk; Lim, Yoon-Soo; Park, Jung-Hwan; Park, Dong-Jun; Yang, Jae-Ho

2018-01-01

Zircaloy-4 is a traditional zirconium-based alloy developed for application in nuclear fuel cladding tubes. The surfaces of Zircaloy-4 tubes were treated using a laser beam to increase their mechanical strength. Laser beam scanning of a tube coated with yttrium oxide (Y2O3) resulted in the formation of a dispersed oxide layer in the tube's surface region. Y2O3 particles penetrated the Zircaloy-4 during the laser treatment and were distributed uniformly in the surface region. The thickness of the dispersed oxide layer varied from 50 to 140 μm depending on the laser beam trajectory. The laser treatment also modified the texture of the tube. The preferred basal orientation along the normal to the tube surface disappeared, and a random structure appeared after laser processing. The most obvious result was an increase in the mechanical strength. The tensile strength of Zircaloy-4 increased by 10-20% with the formation of the dispersed oxide layer. The compressive yield stress also increased, by more than 15%. Brittle fracture was observed in the surface-treated samples during tensile and compressive deformation at room temperature; however, the fracture behavior was changed in ductile at elevated temperatures.

Effects of different binders on microstructure and phase composition of hydroxyapatite Nd-YAG laser clad coatings

NASA Astrophysics Data System (ADS)

Chien, C. S.; Hong, T. F.; Han, T. J.; Kuo, T. Y.; Liao, T. Y.

2011-01-01

The laser clad coating technique can help to produce metallurgical bonding with high bonding strength between the coating layer and the substrate, which has been gradually applied for hydroxyapatite (HA) coating on metallic substrates. In this study, HA powder is mixed with two different binders, namely water glass (WG) and polyvinyl alcohol (PVA), respectively, and is then clad on Ti-6Al-4V substrates using an Nd:YAG laser system under various processing conditions. The microstructure, chemical composition and hardness of the coating layer and transition layer of the various samples are then systematically explored. The experimental results show that the coating layers of the various samples all contain both cellular dendrites and rod-like piled structures, while the transition layers contain only cellular dendrites. For all samples, the coating layer consists mostly of CaTiO 3, Ca 2P 2O 7, CaO and HA phases, whereas the transition layer contains primarily CaTiO 3, Ca 2P 2O 7, Ti 3P, Ti and HA phases. In addition, the transition layer of the WG samples also contains SiO 2 and Si 2Ti phases. In all of the specimens, the transition layer has a higher average hardness than the substrate or coating layer. Moreover, the transition layer in the WG sample is harder than that in the PVA sample.

Anisotropic Azimuthal Power and Temperature distribution on FuelRod. Impact on Hydride Distribution

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

Motta, Arthur; Ivanov, Kostadin; Arramova, Maria

2015-04-29

The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split intomore » two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.« less

2 Professors Rock Out Online to Study Fame--and Us

ERIC Educational Resources Information Center

Young, Jeffrey R.

2009-01-01

Most people who stumble across the YouTube video of the self-proclaimed rock star Gory Bateson singing to a scantily clad prostitute in Amsterdam's red-light district probably have no idea that the work is part of a research project--or that the man holding the guitar is a tenured professor. The video has attracted more than 12,000 views and won a…

Note: Calibration of EBT3 radiochromic film for measuring solar ultraviolet radiation

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

Chun, S. L.; Yu, P. K. N., E-mail: peter.yu@cityu.edu.hk; State Key Laboratory in Marine Pollution, City University of Hong Kong, Kowloon Tong

Solar (UVA + UVB) exposure was assessed using the Gafchromic EBT3 film. The coloration change was represented by the net reflective optical density (Net ROD). Through calibrations against a UV-tube lamp, operational relationships were obtained between Net ROD and the (UVA + UVB) exposures (in J cm⁻²p or J m⁻²). The useful range was from ~0.2 to ~30 J cm⁻². The uniformity of UV irradiation was crucial for an accurate calibration. For solar exposures ranging from 2 to 11 J cm⁻², the predicted Net ROD agreed with the recorded values within 9%, while the predicted exposures agreed with the recordedmore » values within 15%.« less

One hundred angstrom niobium wire

NASA Technical Reports Server (NTRS)

Cline, H. E.; Rose, R. M.; Wulff, J.

1968-01-01

Composite of fine niobium wires in copper is used to study the size and proximity effects of a superconductor in a normal matrix. The niobium rod was drawn to a 100 angstrom diameter wire on a copper tubing.

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

Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

DOE PAGES

Yan, Y.; Qian, S.; Littrell, K.; ...

2015-02-13

A non-destructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentration were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distributionmore » of circumferential hydrides across the wall. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. This study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount ( 20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor will be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for absolute hydrogen concentration determination.« less

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl Alloys

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

Field, Kevin G; Yamamoto, Yukinori; Pint, Bruce A

2016-01-01

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanicalmore » characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs.« less

Investigation of the weldability of iron-aluminum-chromium overlay coatings for corrosion protection in oxidizing/sulfidizing environments

NASA Astrophysics Data System (ADS)

Regina, Jonathan R.

The current study investigated the effect of chromium additions on the hydrogen cracking susceptibility of Fe-Al weld overlay claddings containing chromium additions. It was found that the weldability of FeAlCr claddings was a function of both the aluminum and chromium concentrations of the weld coatings. Weld overlay compositions that were not susceptible to hydrogen cracking were identified and the underlying mechanism behind the hydrogen cracking phenomenon was investigated further. It was concluded that the cracking behavior of the FeAlCr welds depended strongly on the microstructure of the weld fusion zone. Although it was found that the cracking susceptibility was influenced by the presence of Fe-Al intermetallic phases (namely Fe3 Al and FeAl), the cracking behavior of FeAlCr weld overlay claddings also depended on the size and distribution of carbide and oxide particles present within the weld structure. These particles acted as hydrogen trapping sites, which are areas where free hydrogen segregates and can no longer contribute to the hydrogen embrittlement of the metal. It was determined that in practical applications of these FeAlCr weld overlay coatings, carbon should be present within these welds to reduce the amount of hydrogen available for hydrogen cracking. Based on the weldability results of the FeAlCr weld claddings, coating compositions that were able to be deposited crack-free were used for long-term corrosion testing in a simulated low NOx environment. These alloys were compared to a Ni-based superalloy (622), which is commonly utilized as boiler tube coatings in power plant furnaces for corrosion protection. It was found that the FeAlCr alloys demonstrated superior corrosion resistance when compared to the Ni-based superalloy. Due to the excellent long-term corrosion behavior of FeAlCr weld overlays that were immune to hydrogen cracking, it was concluded that select FeAlCr weld overlay compositions would make excellent corrosion resistant coatings for boiler tubes located in low NOx burning environments.

Annular flow diverter valve

DOEpatents

Rider, Robert L.

1980-01-01

A valve for diverting flow from the center of two concentric tubes to the annulus between the tubes or, operating in the reverse direction, for mixing fluids from concentric tubes into a common tube and for controlling the volume ratio of said flow consists of a toroidal baffle disposed in sliding engagement with the interior of the inner tube downstream of a plurality of ports in the inner tube, a plurality of gates in sliding engagement with the interior of the inner tube attached to the baffle for movement therewith, a servomotor having a bullet-shaped plug on the downstream end thereof, and drive rods connecting the servomotor to the toroidal baffle, the servomotor thereby being adapted to move the baffle into mating engagement with the bullet-shaped plug and simultaneously move the gates away from the ports in the inner tube and to move the baffle away from the bullet-shaped plug and simultaneously move the gates to cover the ports in the inner tube.

Temporary arthrodesis using fixator rods in two-stage revision of septic knee prothesis with severe bone and tissue defects.

PubMed

Röhner, Eric; Pfitzner, Tilman; Preininger, Bernd; Zippelius, Timo; Perka, Carsten

2016-01-01

The present study describes a new temporary arthrodesis procedure, which aims for septic knee prosthesis replacement, in particular for larger bone and soft tissue defects. Our technique offers high stability and full weight-bearing capacity of the knee joint. The study included 16 patients with major bone defects (AORI type IIb or greater) after receiving a radical debridement and a septic two-stage revision total knee arthroplasty. After removing the infected prosthesis and debridement, two AO fixator rods were positioned into the intramedullary space of the femur and tibia. Subsequently, both rods were joined tube-to-tube and adjusted in the center of the knee joint. Finally, the whole cavity of the knee joint was filled with PMMA. The number of previous surgeries, bacterial spectrum, risk factors for further infection and reinfection rates was recorded. Immediately after the temporary arthrodesis, radiographs of the knee with the enclosed spacers were taken in order to compare to previous radiographs and avoiding to miss possible spacer loosening. Nine of sixteen patients underwent more than two revision surgeries before receiving our new arthrodesis technique. No cases of spacer loosening were observed in all 16 patients; further, there were no peri-implant fractures, and four persistent infections were noted. Temporary arthrodesis using AO fixator rods offers a high stability without loosening. Its potential to replace conventional augmentation techniques should be taken into account, particularly in the case of larger bone and tissue defects. In clinical practice, the cemented spacer using AO fixator rods could be an alternative technique for temporary knee arthrodesis after septic debridement. Retrospective case series, Level IV.

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.

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

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

Hales, J. D.; Williamson, R. L.; Novascone, S. R.

2016-09-01

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact,more » and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.« less

Nuclear radiation actuated valve

DOEpatents

Christiansen, David W.; Schively, Dixon P.

1985-01-01

A nuclear radiation actuated valve for a nuclear reactor. The valve has a valve first part (such as a valve rod with piston) and a valve second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics. The valve's first part is positioned to receive nuclear radiation from the nuclear reactor's fuel region. The valve's second part is positioned so that its nuclear radiation induced swelling is different from that of the valve's first part. The valve's second part also is positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system.

Proving the viability of manufacturing of multi-layer steel/vanadium alloy/steel composite tubes by numerical simulations and experiment

NASA Astrophysics Data System (ADS)

Nechaykina, T.; Nikulin, S.; Rozhnov, A.; Molotnikov, A.; Zavodchikov, S.; Estrin, Y.

2018-05-01

Vanadium alloys are promising structural materials for fuel cladding tubes for fast-neutron reactors. However, high solubility of oxygen and nitrogen in vanadium alloys at operating temperatures of 700 °C limits their application. In this work, we present a novel composite structure consisting of vanadium alloy V-4Ti-4Cr (provides high long-term strength of the material) and stainless steel Fe-0.2C-13Cr (as a corrosion resistant protective layer). It is produced by co-extrusion of these materials forming a three-layered tube. Finite element simulations were utilised to explore the influence of the various co-extrusion parameters on manufacturability of multi-layered tubes. Experimental verification of the numerical modelling was performed using co-extrusion with the process parameters suggested by the numerical simulations. Scanning electron microscopy and microhardness measurements revealed a defect-free diffusion layer at the interfaces between both materials indicating a good quality bonding for these co-extrusion conditions.

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

Gain-switched laser diode seeded Yb-doped fiber amplifier delivering 11-ps pulses at repetition rates up to 40-MHz

NASA Astrophysics Data System (ADS)

Ryser, Manuel; Neff, Martin; Pilz, Soenke; Burn, Andreas; Romano, Valerio

2012-02-01

Here, we demonstrate all-fiber direct amplification of 11 picosecond pulses from a gain-switched laser diode at 1063 nm. The diode was driven at a repetition rate of 40 MHz and delivered 13 μW of fiber-coupled average output power. For the low output pulse energy of 0.33 pJ we have designed a multi-stage core pumped preamplifier based on single clad Yb-doped fibers in order to keep the contribution of undesired amplified spontaneous emission as low as possible and to minimize temporal and spectral broadening. After the preamplifier we reduced the 40 MHz repetition rate to 1 MHz using a fiber coupled pulse-picker. The final amplification was done with a cladding pumped Yb-doped large mode area fiber and a subsequent Yb-doped rod-type fiber. With our setup we achieved amplification of 72 dBs to an output pulse energy of 5.7 μJ, pulse duration of 11 ps and peak power of >0.6 MW.

Recent advancements in transparent ceramics and crystal fibers for high power lasers

NASA Astrophysics Data System (ADS)

Kim, W.; Baker, C.; Villalobos, G.; Florea, C.; Gibson, D.; Shaw, L. B.; Bowman, S.; Bayya, S.; Sadowski, B.; Hunt, M.; Askins, C.; Peele, J.; Aggarwal, I. D.; Sanghera, J. S.

2013-05-01

In this paper, we present our recent progress in the development of rare-earth (Yb3+ or Ho3+) doped Lu2O3 and Y2O3 sesquioxides for high power solid state lasers. We have fabricated high quality transparent ceramics using nano-powders synthesized by a co-precipitation method. This was accomplished by developments in high purity powder synthesis and low temperature scalable sintering technology developed at NRL. The optical, spectral and morphological properties as well as the lasing performance from our highly transparent ceramics are presented. In the second part of the paper, we discuss our recent research effort in developing cladded-single crystal fibers for high power single frequency fiber lasers has the potential to significantly exceed the capabilities of existing silica fiber based lasers. Single crystal fiber cores with diameters as small as 35μm have been drawn using high purity rare earth doped ceramic or single crystal feed rods by the Laser Heated Pedestal Growth (LHPG) process. Our recent results on the development of suitable claddings on the crystal fiber core are discussed.

Microstructures and Properties of Laser Cladding Al-TiC-CeO2 Composite Coatings

PubMed Central

Kong, Dejun; Song, Renguo

2018-01-01

Al-TiC-CeO2 composite coatings have been prepared by using a laser cladding technique, and the microstructure and properties of the resulting composite coatings have been investigated using scanning electron microscopy (SEM), a 3D microscope system, X-ray diffraction (XRD), micro-hardness testing, X-ray stress measurements, friction and wear testing, and an electrochemical workstation. The results showed that an Al-Fe phase appears in the coatings under different applied laser powers and shows good metallurgical bonding with the matrix. The dilution rate of the coating first decreases and then increases with increasing laser power. The coating was transformed from massive and short rod-like structures into a fine granular structure, and the effect of fine grain strengthening is significant. The microhardness of the coatings first decreases and then increases with increasing laser power, and the maximum microhardness can reach 964.3 HV0.2. In addition, the residual stress of the coating surface was tensile stress, and crack size increases with increasing stress. When the laser power was 1.6 kW, the coating showed high corrosion resistance. PMID:29373555

Microstructures and Properties of Laser Cladding Al-TiC-CeO₂ Composite Coatings.

PubMed

He, Xing; Kong, Dejun; Song, Renguo

2018-01-26

Al-TiC-CeO₂ composite coatings have been prepared by using a laser cladding technique, and the microstructure and properties of the resulting composite coatings have been investigated using scanning electron microscopy (SEM), a 3D microscope system, X-ray diffraction (XRD), micro-hardness testing, X-ray stress measurements, friction and wear testing, and an electrochemical workstation. The results showed that an Al-Fe phase appears in the coatings under different applied laser powers and shows good metallurgical bonding with the matrix. The dilution rate of the coating first decreases and then increases with increasing laser power. The coating was transformed from massive and short rod-like structures into a fine granular structure, and the effect of fine grain strengthening is significant. The microhardness of the coatings first decreases and then increases with increasing laser power, and the maximum microhardness can reach 964.3 HV 0.2 . In addition, the residual stress of the coating surface was tensile stress, and crack size increases with increasing stress. When the laser power was 1.6 kW, the coating showed high corrosion resistance.

Purification of nuclear grade Zr scrap as the high purity dense Zr deposits from Zirlo scrap by electrorefining in LiF-KF-ZrF4 molten fluorides

NASA Astrophysics Data System (ADS)

Park, Kyoung Tae; Lee, Tae Hyuk; Jo, Nam Chan; Nersisyan, Hayk H.; Chun, Byong Sun; Lee, Hyuk Hee; Lee, Jong Hyeon

2013-05-01

Zirconium (Zr) has commonly been used as a cladding material of nuclear fuel. Moreover, it is regarded as the only material that can be used for nuclear fuel cladding because it has the lowest neutron capture cross section of any metal element and because it has high corrosion resistance and size stability. In this study, Hf-free Zr tubes (Zr-1Nb-1Sn-0.1Fe) were used as anode materials and electrorefining was performed in a LiF-KF eutectic 6 wt.% ZrF4 molten fluoride salt system. As a result of electrolysis, Zr scrap metal was recycled into pure Zr with low levels of impurities, and the size and density of the Zr deposit was controlled using applied current density.

Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

NASA Astrophysics Data System (ADS)

Sobolev, V.; Uyttenhove, W.; Thetford, R.; Maschek, W.

2011-07-01

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2-x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of "best estimates" provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

The European scene regarding spallation neutron sources

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

Bauer, G.S.

1996-06-01

In Europe, a short pulse spallation neutron source, ISIS, has been operating for over 10 years, working its way up to a beam power level of 200 kW. A continuous source, SINQ, designed for a beam power of up to 1 MW, is scheduled to start operating at the end of 1996, and a detailed feasibility study has been completed for a 410 kW short pulse source, AUSTRON. Each of these sources seems to have settled for a target concept which is at or near the limits of its feasibility: The ISIS depleted uranium plate targets, heavy water cooled andmore » Zircaloy clad, have so far not shown satisfactory service time and operation is likely to continue with a Ta-plate target, which, in the past has been used successfully for the equivalent of one full-beam-year before it was taken out of service due to degrading thermal properties. SINQ will initially use a rod target, made of Zircaloy only, but plans exist to move on to clad lead rods as quickly as possible. Apart from the not yet explored effect of hydrogen and helium production, there are also concerns about the generation of 7-Be in the cooling water from the spallation of oxygen, which might result in undesirably high radioactivity in the cooling plant room. A Liquid metal target, also under investigation for SINQ, would not only reduce this problem to a level of about 10 %, but would also minimize the risk of radiolytic corrosion in the beam interaction zone. Base on similar arguments, AUSTRON has been designed for edge cooled targets, but thermal and stress analyses show, that this concept is not feasible at higher power levels.« less

The development of fuel performance models at the European institute for transuranium elements

NASA Astrophysics Data System (ADS)

Lassmann, K.; Ronchi, C.; Small, G. J.

1989-07-01

The design and operational performance of fuel rods for nuclear power stations has been the subject of detailed experimental research for over thirty years. In the last two decades the continuous demands for greater economy in conjunction with more stringent safety criteria have led to an increasing reliance on computer simulations. Conditions within a fuel rod must be calculated both for normal operation and for proposed reactor faults. It has thus been necessary to build up a reliable, theoretical understanding of the intricate physical, mechanical and chemical processes occurring under a wide range of conditions to obtain a quantitative insight into the behaviour of the fuel. A prime requirement, which has also proved to be the most taxing, is to predict the conditions under which failure of the cladding might occur, particularly in fuel nearing the end of its useful life. In this paper the general requirements of a fuel performance code are discussed briefly and an account is given of the basic concepts of code construction. An overview is then given of recent progress at the European Institute for Transuranium Elements in the development of a fuel rod performance code for general application and of more detailed mechanistic models for fission product behaviour.

Method and apparatus for automatically tracking a workpiece surface. [Patents

DOEpatents

Not Available

1981-02-03

Laser cutting concepts and apparatus have been developed for cutting the shroud of the core fuel subassemblies. However, much care must be taken in the accuracy of the cutting since the fuel rods within the shroud often become warped and are forced into direct contact with the shroud in random regions. Thus, in order to cut the nuclear fuel rod shroud accurately so as not to puncture the cladding of the fuel rods, and to insure optimal cutting efficiency and performance, the focal point of beam need be maintained accurately at the workpiece surface. It becomes necessary to detect deviations in the level of the workpiece surface accurately in connection with the cutting process. Therefore, a method and apparatus for tracking the surface of a workpiece being cut by a laser beam coming from a focus head assembly is disclosed which includes two collimated laser beams directed onto the work-piece surface at spaced points by beam directing optics in generally parallel planes of incidence. A shift in spacing between the two points is detected by means of a video camera system and processed by a computer to yield a workpiece surface displacement signal which is input to a motor which raises or lowers the beam focus head accordingly.

Methodology for Mechanical Property Testing on Fuel Cladding Using an Expanded Plug Wedge Test

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

Wang, Jy-An John; Jiang, Hao

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at ORNL and is described fully in US Patent Application 20060070455, Expanded plug method for developing circumferential mechanical properties of tubular materials. This method is designed for testing fuel rod cladding ductility in a hot cell utilizing an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the testmore » component assembly in the hot cell and the direct measurement of specimen strain. It was also found that cladding strength could be determined from the test results. The basic approach of this test method is to apply an axial compressive load to a cylindrical plug of polyurethane (or other materials) fitted inside a short ring of the test material to achieve radial expansion of the specimen. The diameter increase of the specimen is used to calculate the circumferential strain accrued during the test. The other two basic measurements are total applied load and amount of plug compression (extension). A simple procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves. However, several deficiencies exist in this expanded-plug loading ring test, which will impact accuracy of test results and introduce potential shear failure of the specimen due to inherited large axial compressive stress from the expansion plug test. First of all, the highly non-uniform stress and strain distribution resulted in the gage section of the clad. To ensure reliable testing and test repeatability, the potential for highly non-uniform stress distribution or displacement/strain deformation has to be eliminated at the gage section of the specimen. Second, significant compressive stresses were induced by clad bending deformation due to a clad bulging effect (or the barreling effect). The barreling effect caused very large localized shear stress in the clad and left testing material at a high risk of shear failure. The above combined effects will result in highly non-conservative predictions both in strength and ductility of the tested clad, and the associated mechanical properties as well. To overcome/mitigate the mentioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. Through detailed parameter investigation on specific geometry designs, careful filtering of material for the expansion plug, as well as adding newly designed parts to the testing system, a method to reconcile the potential non-conservatism embedded in the expansion plug test system has been discovered. A modified expansion plug testing protocol has been developed based on the method. In order to closely resemble thin-wall theory, a general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor called -factor is defined to correlate the ring load P into hoop stress . , = . The generated stress-strain curve agrees very well with tensile test data in both the elastic and plastic regions.« 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.

Synthesis of barium-strontium titanate hollow tubes using Kirkendall effect

NASA Astrophysics Data System (ADS)

Chen, Xuncai; Im, SangHyuk; Kim, Jinsoo; Kim, Woo-Sik

2018-02-01

(BaSr)TiO3 hexagonal hollow tubes was fabricated by a solid-state interfacial reaction including a Kirkendall diffusion. Using a co-precipitation and sol-gel process, a core@shell structure of (BaSr)CO3@TiO2 rods were prepared, and then converted to (BaSr)TiO3 hollow tubes at 750 °C. This was a first achievement of single-phase crystal hollow tube. Here, the inner diameter and wall thickness of hollow tube were about 700 nm and 130 nm, respectively. The fabrication of (BaSr)TiO3 hollow tubes was monitored with scanning electron microscopy (SEM), energy-dispersive spectrometry (EDS), transmission electron microscopy (TEM), and X-ray diffraction (XRD) to investigate their formation mechanism. The present synthetic approach would provide a new insight into the design and fabrication of hollow architectures of many perovskite oxides.

Process development of two high strength tantalum base alloys (ASTAR-1211C and ASTAR-1511C)

NASA Technical Reports Server (NTRS)

Ammon, R. L.

1974-01-01

Two tantalum base alloys, Ta-12W-1.0Re-0.7Hf-0.025C(ASTAR-1211C) and Ta-15W-1.0Re-0.7Hf-0.025C(ASTAR-1511C), were cast as 12.5 cm (5 inch) diameter ingots and processed to swaged rod, sheet, forged plate, and tubing. Swaged rod was evaluated with respect to low temperature ductility, elevated temperature tensile properties, and elevated temperature creep behavior. A standard swaging process and final annealing schedule were determined. Elevated temperature tensile properties, low temperature impact properties, low temperature DBTT behavior, and extended elevated temperature creep properties were determined. A process for producing ASTAR-1211C and ASTAR-1511C sheet were developed. The DBTT properties of GTA and EB weld sheet given post-weld anneal and thermal aging treatments were determined using bend and tensile specimens. High and low temperature mechanical properties of forging ASTAR-1211C and ASTAR-1511C plate were determined as well as elevated temperature creep properties. Attempts to produce ASTAR-1211C tubing were partially successful while attempts to make ASTAR-1511C tubing were completely unsuccessful.

Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

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

Ernst, Frank

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, andmore » fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.« less

Apparatus facilitates high-temperature tensile testing in vacuum

NASA Technical Reports Server (NTRS)

Sikora, P. F.

1964-01-01

An apparutus for heating refractory materials to high temperatures during tensile testing includes a water-cooled stainless steel vacuum chamber. This contains a resistance heater consisting of a slit tube of tantalum or tungsten to enclose the tensile test rod.

R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

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

Szilard, Ronaldo Henriques; Zhang, Hongbin; Epiney, Aaron Simon

The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licenseemore » costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize plant safety and operational margins. Advanced MOOSE tools that are needed to complete this integrated evaluation model are: RAVEN, RELAP-7, BISON, and MAMMOTH.« less

Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

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

Not Available

1978-01-01

This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified.

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.

Recent progress in online ultrasonic process monitoring

NASA Astrophysics Data System (ADS)

Wen, Szu-Sheng L.; Chen, Tzu-Fang; Ramos-Franca, Demartonne; Nguyen, Ky T.; Jen, Cheng-Kuei; Ihara, Ikuo; Derdouri, A.; Garcia-Rejon, Andres

1998-03-01

On-line ultrasonic monitoring of polymer co-extrusion and gas-assisted injection molding are presented. During the co- extrusion of high density polyethylene and Santoprene ultrasonic sensors consisting of piezoelectric transducers and clad ultrasonic buffer rods are used to detect the interface between these two polymers and the stability of the extrusion. The same ultrasonic sensor also measures the surface temperature of the extruded polymer. The results indicate that temperature measurements using ultrasound have a faster response time than those obtained by conventional thermocouple. In gas-assisted injection molding the polymer and gas flow front positions are monitored simultaneously. This information may be used to control the plunger movement.

METHOD OF MAKING WIRE FUEL ELEMENTS

DOEpatents

Zambrow, J.L.

1960-08-01

A method is given for making a nuclear reactor fuel element in the form of a uranium-bearing wire clad with zirconium. A uranium bar is enclosed in a zirconium sheath which is coated with an oxide of magnesium, beryllium, or zirconium. The sheathed bar is then placed in a steel tube and reduced to the desired diameter by swaging at 800 to 900 deg C, after which the steel and oxide are removed.

A novel cluster-tube self-adaptive robot hand.

PubMed

Fu, Hong; Yang, Haokun; Song, Weishu; Zhang, Wenzeng

2017-01-01

This paper proposes a novel cluster-tube self-adaptive robot hand (CTSA Hand). The CTSA Hand consists of a base, a motor, a transmission mechanism, multiple elastic tendons, and a group of sliding-tube assemblies. Each sliding-tube assembly is composed of a sliding tube, a guide rod, two springs and a hinge. When the hand grasping an object, the object pushes some sliding tubes to different positions according to the surface shape of the object, the motor pulls the tendons tight to cluster tubes. The CTSA Hand can realize self-adaptive grasping of objects of different sizes and shapes. The CTSA Hand can grasp multiple objects simultaneously because the grasping of the hand acts as many grippers in different directions and heights. The grasping forces of the hand are adjusted by a closed-loop control system with potentiometer. Experimental results show that the CTSA Hand has the features of highly self-adaption and large grasping forces when grasping various objects.

NDE of cylindrically symmetric components with piezofilm transducers

NASA Astrophysics Data System (ADS)

Hsu, David K.; Zhang, Zhong

PVDF polymer film transducers are presently shown to exhibit the flexibility and comformability required for inspection of components with curved surfaces. Although these transducers are less efficient than rigid ceramic ones, and are less accurately matched to the acoustic impedance of metals as well as ceramic transducers, their advantages are presently shown to outweigh their disadvantages in some applications involving tube and rod shaped components. Interface measurements of a Zr/Zircalloy-2 tube allowed the detailed evaluation of weakly reflecting interfaces.

ONR Workshop on Magnetohydrodynamic Submarine Propulsion (2nd), Held in San Diego, California on November 16-17, 1989

DTIC Science & Technology

1990-07-01

electrohtic dissociation of the electrode mate- pedo applications seem to be still somewhat rial, and to provide a good gas evolution wlhich out of the...rod cathode. A unique feature of this preliminary experiment was the use of a prototype gated, intensified video camera. This camera is based on a...microprocessor controlled microchannel plate intensifier tube. The intensifier tube image is focused on a standard CCD video camera so that the object

Multileg Heat-Pipe Evaporator

NASA Technical Reports Server (NTRS)

Alario, J. P.; Haslett, R. A.

1986-01-01

Parallel pipes provide high heat flow from small heat exchanger. Six parallel heat pipes extract heat from overlying heat exchanger, forming evaporator. Vapor channel in pipe contains wick that extends into screen tube in liquid channel. Rods in each channel hold wick and screen tube in place. Evaporator compact rather than extended and more compatible with existing heat-exchanger geometries. Prototype six-pipe evaporator only 0.3 m wide and 0.71 m long. With ammonia as working fluid, transports heat to finned condenser at rate of 1,200 W.

40 CFR 420.111 - Specialized definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... steel products such as coiled wire, rods, and tubes in discrete batches or bundles. (b) The term continuous means those alkaline cleaning operations which process steel products other than in discrete... AND STANDARDS IRON AND STEEL MANUFACTURING POINT SOURCE CATEGORY Alkaline Cleaning Subcategory § 420...

Magnetic field sensor based on the magnetic-fluid-clad combined with singlemode-multimode-singlemode fiber and large core-offset splicing structure

NASA Astrophysics Data System (ADS)

Lv, Ri-qing; Qian, Jun-kai; Zhao, Yong

2018-03-01

A simple, compact optical fiber magnetic field sensor is proposed and experimentally demonstrated in this paper. It is based on the magnetic-fluid-clad combined with singlemode-multimode-singlemode fiber structure and large core-offset splicing structure. It was protected by a section of capillary tube and was sealed by UV glue. A sensing property study of the combined optical fiber structure and the proposed sensor were carried out. The experimental results show that the sensitivity of the refractive index of the optical fiber sensing structure is up to 156.63 nm/RIU and the magnetic field sensitivity of the proposed sensor is up to -97.24 pm/Oe in the range from 72.4 Oe to 297.8 Oe. The proposed sensor has several other advantages, such as simple structure, small size, easy fabrication and low cost.

Performance of OSC's initial Amtec generator design, and comparison with JPL's Europa Orbiter goals

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

Schock, A.; Noravian, H.; Or, C.

1998-07-01

The procedure for the analysis (with overpotential correction) of multitube AMTEC (Alkali Metal Thermal-to-Electrical Conversion) cells described in Paper IECEC 98-243 was applied to a wide range of multicell radioisotope space power systems. System design options consisting of one or two generators, each with 2, 3, or 4 stacked GPHS (General Purpose Heat Source) modules, identical to those used on previous NASA missions, were analyzed and performance-mapped. The initial generators analyzed by OSC had 8 AMTEC cells on each end of the heat source stack, with five beta-alumina solid electrolyte (BASE) tubes per cell. The heat source and converters inmore » the Orbital generator designs are embedded in a thermal insulation system consisting of Min-K fibrous insulation surrounded by graded-length molybdenum multifoils. Detailed analyses in previous Orbital studies found that such an insulation system could reduce extraneous heat losses to about 10%. For the above design options, the present paper presents the system mass and performance (i.e., the EOM system efficiency and power output and the BOM evaporator and clad temperatures) for a wide range of heat inputs and load voltages, and compares the results with JPL's preliminary goals for the Europa Orbiter mission to be launched in November 2003. The analytical results showed that the initial 16-cell generator designs resulted in either excessive evaporator and clad temperatures and/or insufficient power outputs to meet the JPL-specified mission goals. The computed performance of modified OSC generators with different numbers of AMTEC cells, cell diameters, cell lengths, cell materials, BASE tube lengths, and number of tubes per cell are described in Paper IECEC.98.245 in these proceedings.« less

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

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

Linton, Kory D.; Field, Kevin G.; Petrie, Christian M.

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. Tomore » address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).« less

40 CFR 63.1156 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Standards for Hazardous Air Pollutants for Steel Pickling-HCl Process Facilities and Hydrochloric Acid... to remove residual acid. Carbon steel means steel that contains approximately 2 percent or less... equipment and tanks configured for pickling metal strip, rod, wire, tube, or pipe that is passed through an...

40 CFR 63.1156 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Standards for Hazardous Air Pollutants for Steel Pickling-HCl Process Facilities and Hydrochloric Acid... to remove residual acid. Carbon steel means steel that contains approximately 2 percent or less... equipment and tanks configured for pickling metal strip, rod, wire, tube, or pipe that is passed through an...

A Simple Experiment to Explore Standing Waves in a Flexible Corrugated Sound Tube

NASA Astrophysics Data System (ADS)

Amorim, Maria Eva; Sousa, Teresa Delmira; Carvalho, P. Simeão; Sousa, Adriano Sampaioe

2011-09-01

Sound tubes, pipes, and singing rods are used as musical instruments and as toys to perform amusing experiments. In particular, corrugated tubes present unique characteristics with respect to the sounds they can produce; that is why they have been studied so intensively, both at theoretical and experimental levels.1-4 Experimental studies usually involve expensive and sophisticated equipment that is out of reach of school laboratory facilities.3-6 In this paper we show how to investigate quantitatively the sounds produced by a flexible sound tube corrugated on the inside by using educational equipment readily available in school laboratories, such as the oscilloscope, the microphone, the anemometer, and the air pump. We show that it is possible for students to study the discontinuous spectrum of sounds produced by a flexible corrugated tube and go even further, computing the speed of sound in air with a simple experimental procedure.

Crippling Strength of Axially Loaded Rods

NASA Technical Reports Server (NTRS)

Natalis, FR

1921-01-01

A new empirical formula was developed that holds good for any length and any material of a rod, and agrees well with the results of extensive strength tests. To facilitate calculations, three tables are included, giving the crippling load for solid and hollow sectioned wooden rods of different thickness and length, as well as for steel tubes manufactured according to the standards of Army Air Services Inspection. Further, a graphical method of calculation of the breaking load is derived in which a single curve is employed for determination of the allowable fiber stress. Finally, the theory is discussed of the elastic curve for a rod subject to compression, according to which no deflection occurs, and the apparent contradiction of this conclusion by test results is attributed to the fact that the rods under test are not perfectly straight, or that the wall thickness and the material are not uniform. Under the assumption of an eccentric rod having a slight initial bend according to a sine curve, a simple formula for the deflection is derived, which shows a surprising agreement with test results. From this a further formula is derived for the determination of the allowable load on an eccentric rod. The resulting relations are made clearer by means of a graphical representation of the relation of the moments of the outer and inner forces to the deflection.

Test Results of Heat Exchanger Cleaning in Support of Ocean Thermal Energy Conversion.

DTIC Science & Technology

1980-12-01

tests evaluated the performance of three in-situ cleaning techniques in two potential heat exchanger materials ...1-6. 41Mann, M. J., 1979, "Possible Cu-Ni-Clad Steel Material and Abrasive Slurry Cleaning System for Plate-Fin-Type OTEC Heat Exchangers ," in...of a Shell-less Folded Aluminum Tube, OTEC Heat Exchanger ," Proceedings of the Sixth OTEC Conference, Washington, DC, June 19-22, 1978, pp 12.8-1

Nuclear reactor with low-level core coolant intake

DOEpatents

Challberg, Roy C.; Townsend, Harold E.

1993-01-01

A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

Qualification and characterization of electronics of the fast neutron Hodoscope detectors using neutrons from CABRI core

NASA Astrophysics Data System (ADS)

Mirotta, S.; Guillot, J.; Chevalier, V.; Biard, B.

2018-01-01

The study of Reactivity Initiated Accidents (RIA) is important to determine up to which limits nuclear fuels can withstand such accidents without clad failure. The CABRI International Program (CIP), conducted by IRSN under an OECD/NEA agreement, has been launched to perform representative RIA Integral Effect Tests (IET) on real irradiated fuel rods in prototypical Pressurized Water Reactors (PWR) conditions. For this purpose, the CABRI experimental pulse reactor, operated by CEA in Cadarache, France, has been strongly renovated, and equipped with a pressurized water loop. The behavior of the test rod, located in that loop in the center of the driver core, is followed in real time during the power transients thanks to the hodoscope, a unique online fuel motion monitoring system, and one of the major distinctive features of CABRI. The hodoscope measures the fast neutrons emitted by the tested rod during the power pulse with a complete set of 153 Fission Chambers and 153 Proton Recoil Counters. During the CABRI facility renovation, the electronic chain of these detectors has been upgraded. In this paper, the performance of the new system is presented describing gain calibration methodology in order to get maximal Signal/Noise ratio for amplification modules, threshold tuning methodology for the discrimination modules (old and new ones), and linear detectors response limit versus different reactor powers for the whole electronic chain.

YouTube and the single-rod contraceptive implant: a content analysis.

PubMed

Paul, Jennifer; Boraas, Christy M; Duvet, Mildred; Chang, Judy C

2017-07-01

Since the internet has become an important source of contraceptive information with YouTube.com as the second most visited site, we analysed contraceptive implant YouTube videos for content and clinical accuracy. Using the terms 'contraceptive implant', 'Nexplanon' and 'Implanon', the top 20 results on YouTube by relevance and view count were identified. After excluding duplicates, single-rod implant videos in English were included. Videos were classified as providing a professional or patient perspective. Views, duration and comments were noted. Videos were rated for reliability, global quality scale and whether they were positive or negative about the implant. Inter-rater agreement was measured. A total of 120 videos were retrieved; 52 were eligible for review. Less than 23% were professional videos; the majority reported patient experience (46% testimonials, 27% real-time procedure videos, 4% other). Patient videos had been posted a significantly longer duration of time than professional videos (364 vs 188 days, p =0.02), were less reliable ( p ≤0.0001) and were of lower global quality ( p <0.0001). Some 61% of implant testimonial videos were rated as 'positive experiences' and inter-rater agreement was very good (κ=0.81). All testimonials mentioned side effects, commonly irregular bleeding and discomfort with insertion. A minority (26%) reported misinformation. This study found that most of the information on YouTube pertaining to contraceptive implants is accurate, is presented from the patient's perspective, and promotes the method's use. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

TEM00 mode Nd:YAG solar laser by side-pumping a grooved rod

NASA Astrophysics Data System (ADS)

Vistas, Cláudia R.; Liang, Dawei; Almeida, Joana; Guillot, Emmanuel

2016-05-01

A simple TEM00 mode solar laser system with a grooved Nd:YAG rod pumped through a heliostat-parabolic mirror system is reported here. The radiation coupling capacity of a fused silica tube lens was combined with the multipass pumping ability of a 2 V-shaped cavity to provide efficient side-pumping along a 4.0 mm diameter grooved Nd:YAG single-crystal rod. TEM00 mode solar laser power of 3.4 W was measured by adopting an asymmetric large-mode laser resonant cavity. Record TEM00 mode solar laser collection efficiency of 3.4 W/m2and slope efficiency of 1.9% was achieved, which corresponds to 1.8 and 2.4 times more than the previous TEM00 mode Nd:YAG solar laser using the PROMES-CNRS heliostat-parabolic mirror system, respectively.

Pretest analysis document for Semiscale Test S-LH-1

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

Shaw, R.A.

Results from various pretest calculations which were performed for Test S-LH-1 are included in this report. Test S-LH-1 has been designed to produce primary liquid holdup in the steam generator U-tubes similar to Tests S-UT-8. The analyses included in this report indicate liquid will be held in the tubes, the core liquid level will be appropriately depressed, and a core heater rod temperature excursion should occur. Several sensitivity studies are also included which identify parameters which could affect the response.

Erbium Distribution in Single Crystal YAG Fibers Grown by Laser-Heated Pedestal Growth Technique

DTIC Science & Technology

2015-08-28

single crystal YAG fibers grown by laser - heated pedestal growth technique Single crystal (SC) yttrium aluminum garnet (YAG, Y3Al5O12) as a host...inserted into a SC YAG tube. This rod-in-tube was used as a preform in our laser -heated pedestal growth (LHPG) apparatus to grow a fiber with a radial...fibers grown by laser -heated pedestal growth technique Report Title Single crystal (SC) yttrium aluminum garnet (YAG, Y3Al5O12) as a host material has

METHOD AND APPARATUS FOR FABRICATING TUBULAR UNITS

DOEpatents

Haldeman, G.W.

1959-02-24

A method and apparatus are described for fabricating tubular assemblies such as clad fuel elements for nuclear reactors. According to this method, a plurality of relatively short cylindrical slug-shaped members are inserted in an outer protective tubular jacket, and the assembly is passed through a reducing die to draw the outer tubular member into tight contact with the slug members, the slugs being automatically spaced with respect to each other and helium being inserted during the drawing operation to fill the spaces. The apparatus includes a pusher rod which functions to space the slugelements equidistantly by pushing on them in the direction of drawing but traveling at a slower rate than that of the tubular member.

Tubular gage for a liquid-metal-cooled fast breeder reactor

DOEpatents

Hutter, Ernest; Tuma, Leroy A.

1977-06-14

Spring-loaded plungers are arranged about a housing for insertion into a polygonal tube, one plunger for each side of the tube. Each plunger has a locking cam and sliding wedge mechanism which can overcome the spring force associated with the plunger and lock it in any position. The wedges are operated by a rod moveable axially in the housing. Several housings with their associated plungers can be stacked. The stack is lowered into the polygonal tube with all of the plungers locked in a fully inward position. When the stack is in the tube, each wedge is moved to release its locking cam, allowing each of the plungers to spring outward against an inner side of the tube. Each housing will thus gage the internal dimensions of the tube at its elevation. The plungers are locked in position, the entire stack is rotated to bring the plungers into the corners described by the intersections of the flat sides, and the stack is removed from the tube whereupon the dimensions across opposite locked plungers may be read by a micrometer.

Fabrication and processing of high-strength densely packed carbon nanotube yarns without solution processes.

PubMed

Liu, Kai; Zhu, Feng; Liu, Liang; Sun, Yinghui; Fan, Shoushan; Jiang, Kaili

2012-06-07

Defects of carbon nanotubes, weak tube-tube interactions, and weak carbon nanotube joints are bottlenecks for obtaining high-strength carbon nanotube yarns. Some solution processes are usually required to overcome these drawbacks. Here we fabricate ultra-long and densely packed pure carbon nanotube yarns by a two-rotator twisting setup with the aid of some tensioning rods. The densely packed structure enhances the tube-tube interactions, thus making high tensile strengths of carbon nanotube yarns up to 1.6 GPa. We further use a sweeping laser to thermally treat as-produced yarns for recovering defects of carbon nanotubes and possibly welding carbon nanotube joints, which improves their Young's modulus by up to ∼70%. The spinning and laser sweeping processes are solution-free and capable of being assembled together to produce high-strength yarns continuously as desired.

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.

Developpement d'un systeme pour la mesure du taux de vide dans un ecoulement diphasique par une methode utilisant des micro-ondes

NASA Astrophysics Data System (ADS)

Pochet, Steven

The measurement of the void fraction is an important parameter in many industrial fields. Whether it is to prevent the phenomenon of critical heat flux in heat tube of thermal power plants, the explosion of gas pockets in oil rigs’ pipes or to detect bubbles in medical catheters, the knowledge of the void fraction can be a key parameter in many diverse applications. Several invasive and non-invasive measurements techniques have been developed these last decades and are based on the difference between the physical properties of liquid and gas. Some of these techniques are not always possible to implement due to restrictions in the geometry of tubes or regulatory standards limiting their use. Throughout this work we propose a new non-invasive void fraction measurement technique based on the reflection of electromagnetic waves on the water-air interface of the mixture. The reflection of electromagnetic wave is induced by a change in the impedance of the propagation medium. The impedance is function of the dielectric properties of the medium. The characteristics of air and water being distinct, it is possible to calculate the complex reflection coefficient at the interface of a double phase mixture. To this end, mathematical modeling of the response of an electromagnetic wave in a tube containing a two phase mixture was made using the model of transmission lines, applicable to microwave frequencies we use. The effects of the amount of air in water and the position of the bubbles in the section of the tube were simulated. It was shown that the phase of the reflected wave was sensitive to the position of bubbles in the tube’s section and that the magnitude of the reflection coefficient varied with the mixture’s void fraction. Subsequently, we designed and built a six-ports reflectometer operating at 2.45 GHz. This system allows the processing and calculation of the reflected wave from the incident wave. A six-ports network, a patch antenna, a wave generator and an amplifier were simulated using HFSS and ADS software. They were then built using the technology of micro-strips on dielectric laminates and the entire system was then calibrated at 6 different frequencies near 2.45 GHz. To this end, we used 4 and 5 loads calibration algorithms that gave us calibrated results with less than 2 % errors. Afterwards, the system was implemented: the antenna was placed tangent to the wall of a vertical tube and connected to the six-ports which was connected to a computer recording and displaying the results in real time. A valve positioned under the tube allows air into the tube and to vary the flow rate. The results showed that the system was sensitive to changes in void fraction from 1% and followed the predictions of the simulated model to a void fraction of about 10%. Possibly du to a change in the structure of the flow for a void fraction of 10%, the signal no longer varies monotonically with respect to the increasing void fraction possibly because of a change in the flow’s configuration. It was shown that the Rayleigh scattering phenomena of air bubbles was involved in the reflection coefficient response. Pictures of the stream at various void fraction state were taken and confirmed a change in the flow’s configuration. By placing a Plexiglas rod to simulate a flow geometry located in the section of the tube, it was noted that the change in phase of the reflected wave was the same as the model when the rod was placed in an empty tube (very few attenuation loss environment). Hence, it is possible to determine the distance of an object in a section of tube from the measurement of the reflected wave’s phase. When the rod is in a very absorbent medium such as water, it is possible to detect a moving rod when it is sufficiently close to the antenna (less than two wavelengths) thanks again to the phase variation. However, detection is still much more difficult due to the absorption of water and can not function effectively for tubes with high diameters compared with the electromagnetic wavelength used.

77 FR 42454 - Airworthiness Directives; Piaggio Aero Industries S.p.A.

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-19

... the brake hydraulic fluid from leaking because of the brake assembly rods contacting the brake valve... session during which conflicting inputs were given to the brake pads between pilot and copilot, a brake... tubings connected to the brake valves, with consequent fluid leakage. Prompted by these findings, PAI...

78 FR 58978 - Airworthiness Directives; Airbus Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-25

... during low speed taxi maneuvers. The bogie fractured aft of the pivot point and remained attached to the sliding tube by the brake torque reaction rods. After this RH bogie failure, the aeroplane continued for... requirements.'' Under that section, Congress charges the FAA with promoting safe flight of civil aircraft in...

Consolidation and fabrication techniques for vanadium-20 w/o titanium /TV-20/

NASA Technical Reports Server (NTRS)

Burt, W. R.; Karasek, F. J.; Kramer, W. C.; Mayfield, R. M.; Mc Gowan, R. D.

1968-01-01

Tests of the mechanical properties, fuel compatibility, sodium corrosion and irradiation behavior were made for vanadium and vanadium alloy. Improved methods for consolidation and fabrication of bar, rod, sheet, and high-quality, small diameter, thin-wall tubing of vanadium-20 without titanium are reported.

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT ...

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

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT BREAKS IN CLADDING OF FUEL ELEMENTS. COUNT-RATE METER IN TOP PANEL INDICATES AMOUNT OF RADIOACTIVITY. LOWER PANELS SUPPLY POWER AND AMPLIFICATION OF SIGNALS GENERATED BY SCINTILLATION COUNTER/PHOTOMULTIPLIER TUBE COMBINATION IN RESPONSE TO RADIOACTIVITY IN A SAMPLE OF THE COOLING WATER. INL NEGATIVE NO. 56-771. Jack L. Anderson, Photographer, 3/15/1956. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

An Inexpensive and Versatile Version of Kundt's Tube for Measuring the Speed of Sound in Air

NASA Astrophysics Data System (ADS)

Papacosta, Pangratios; Linscheid, Nathan

2016-01-01

Experiments that measure the speed of sound in air are common in high schools and colleges. In the Kundt's tube experiment, a horizontal air column is adjusted until a resonance mode is achieved for a specific frequency of sound. When this happens, the cork dust in the tube is disturbed at the displacement antinode regions. The location of the displacement antinodes enables the measurement of the wavelength of the sound that is being used. This paper describes a design that uses a speaker instead of the traditional aluminum rod as the sound source. This allows the use of multiple sound frequencies that yield a much more accurate speed of sound in air.

Analysis for Material Behavior of Sabot/Rods During Launch by Finite Element Method

NASA Astrophysics Data System (ADS)

Kim, Jin Bong; Kim, Man Geun

This study has been investigated to predict the deformation and states of stress and strain by axial and lateral acceleration during launch. Because a gun tube is not perfectly straight at its initial state while under gravity loading, the projectile deforms due to the change of contacts points with the flexible gun tube. Numerical simulations were used for gravity loading and the other type is initial shape and gravity loading. The ANSYS engineering analysis code was used to generate a parametric model of the projectile and conduct finite element analyses. Four types of nonlinear material and contact elements were incorporated into the model to account for the plastic deformation and contact between the penetrator, sabot, and tube.

Generation of flash x-rays using a mercury-anode radiation tube

NASA Astrophysics Data System (ADS)

Oizumi, Teiji; Sato, Eiichi; Sagae, Michiaki; Hayasi, Yasuomi; Tamakawa, Yoshiharu; Yanagisawa, Toru

1993-02-01

The constructions and the radiographic characteristics of a flash x-ray generator having a liquid-anode radiation tube are described. This generator consisted of the following essential components: a high-voltage power supply, a combined ceramic condenser of 10.7 nF, an oil- diffusion pump, an oil circulator, a trigger device, and a flash x-ray tube. The x-ray tube was of a triode and was composed of the following major devices: a mercury anode, a rod-shaped graphite cathode, a trigger electrode made from a copper wire, an x-ray window made from a polyethyleneterephthalate film, and a glass tube body. The ceramic condenser was charged from 40 to 60 kV by a power supply, and the electric charges in the condenser were discharged to the x-ray tube after the triggering. The maximum tube voltage was equivalent to the initial charged voltage of the condenser, and the tube current was less than 0.7 kA. The pulse widths of the flash x rays had values of about 1 microsecond(s) , and the time-integrated x-ray intensity was about 2.4 (mu) C/kg at 0.26 m per pulse with a charged voltage of 60 kV.

Rewetting of hot vertical rod during jet impingement surface cooling

NASA Astrophysics Data System (ADS)

Agrawal, Chitranjan; Kumar, Ravi; Gupta, Akhilesh; Chatterjee, Barun

2016-06-01

A stainless steel (SS-316) vertical rod of 12 mm diameter at 800 ± 10 °C initial temperature was cooled by normal impinging round water jet. The surface rewetting phenomenon was investigated for a range of jet diameter 2.5-4.8 mm and jet Reynolds number 5000-24,000 using a straight tube type nozzle. The investigation were made from the stagnation point to maximum 40 mm downstream locations, simultaneously for both upside and downside directions. The cooling performance of the vertical rod was evaluated on the basis of rewetting parameters i.e. rewetting temperature, wetting delay, rewetting velocity and the maximum surface heat flux. Two separate Correlations have been proposed for the dimensionless rewetting velocity in terms of rewetting number and the maximum surface heat flux that predicts the experimental data within an error band of ±20 and ±15 % respectively.

Nuclear reactor safety device

DOEpatents

Hutter, Ernest

1986-01-01

A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

Multiphoton laser ionization for energy conversion in barium vapor

NASA Astrophysics Data System (ADS)

Makdisi, Y.; Kokaj, J.; Afrousheh, K.; Mathew, J.; Nair, R.; Pichler, G.

2013-03-01

We have studied the ion detection of barium atoms in special heated ovens with a tungsten rod in the middle of the stainless steel tube. The tungsten rod was heated indirectly by the oven body heaters. A bias voltage between the cell body and the tungsten rod of 9 V was used to collect electrons, after the barium ions had been created. However, we could collect the electrons even without the bias voltage, although with ten times less efficiency. We studied the conditions for the successful bias-less thermionic signal detection using excimer/dye laser two-photon excitation of Rydberg states below and above the first ionization limit (two-photon wavelength at 475.79 nm). We employed a hot-pipe oven and heat-pipe oven (with inserted mesh) in order to generate different barium vapor distributions inside the oven. The thermionic signal increased by a factor of two under heat-pipe oven conditions.

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

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

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Experimental study on heat transfer to supercritical water flowing through tubes

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

Zhao, M.; Gu, H.; Cheng, X.

2012-07-01

A test facility named SWAMUP (Supercritical Water Multi-Purpose Loop) has been constructed in Shanghai Jiao Tong Univ. to investigate heat transfer and pressure drop through tubes and rod bundles. SWAMUP is a closed loop with operating pressure up to 30 MPa, outlet-water temperature up to 550 deg. C, and mass flow rate up to 5 t/h. In this paper, experimental study has been carried out on heat transfer of supercritical water flowing vertically through tubes (ID=7.6 and 10 mm). A large number of test points in tubes has been obtained with a wide range of heat flux (200-1500 kw/m{sup 2})more » and mass flux (450-2000 kg/m{sup 2}s). Test results showed that heat transfer deterioration (HTD) caused by buoyancy effect only appears in upward flow and HTD caused by acceleration effect appears both in upward flow and downward flow. The heat transfer coefficients (HTC) produced in tube tests were compared with existing heat transfer correlations. (authors)« less

Condensation enhancement by means of electrohydrodynamic techniques

NASA Astrophysics Data System (ADS)

Butrymowicz, Dariusz; Karwacki, Jarosław; Trela, Marian

2014-12-01

Short state-of-the-art on the enhancement of condensation heat transfer techniques by means of condensate drainage is presented in this paper. The electrohydrodynamic (EHD) technique is suitable for dielectric media used in refrigeration, organic Rankine cycles and heat pump devices. The electric field is commonly generated in the case of horizontal tubes by means of a rod-type electrode or mesh electrodes. Authors proposed two geometries in the presented own experimental investigations. The first one was an electrode placed just beneath the tube bottom and the second one consisted of a horizontal finned tube with a double electrode placed beneath the tube. The experimental investigations of these two configurations for condensation of refrigerant R-123 have been accomplished. The obtained results confirmed that the application of the EHD technique for the investigated tube and electrode arrangement caused significant increase in heat transfer coefficient. The condensation enhancement depends both on the geometry of the electrode system and on the applied voltage.

26 CFR 48.4061(b)-3 - Rebuilt, reconditioned, or repaired parts or accessories.

Code of Federal Regulations, 2010 CFR

2010-04-01

..., Tires, Tubes, Tread Rubber, and Taxable Fuel Automotive and Related Items § 48.4061(b)-3 Rebuilt... batteries, (2) rebabbited or machined connecting rods, (3) reassembled clutches after operations such as the... reassembling (with any necessary replacements of worn parts) of automobile parts or accessories, such as fuel...

10 CFR 31.12 - General license for certain items and self-luminous products containing radium-226.

Code of Federal Regulations, 2014 CFR

2014-01-01

... spinthariscopes), electron tubes, lightning rods, ionization sources, static eliminators, or as designated by the..., or transfer, in accordance with the provisions of paragraphs (b), (c), and (d) of this section, radium-226 contained in the following products manufactured prior to November 30, 2007. (1) Antiquities...

10 CFR 31.12 - General license for certain items and self-luminous products containing radium-226.

Code of Federal Regulations, 2013 CFR

2013-01-01

... spinthariscopes), electron tubes, lightning rods, ionization sources, static eliminators, or as designated by the..., or transfer, in accordance with the provisions of paragraphs (b), (c), and (d) of this section, radium-226 contained in the following products manufactured prior to November 30, 2007. (1) Antiquities...

10 CFR 31.12 - General license for certain items and self-luminous products containing radium-226.

Code of Federal Regulations, 2012 CFR

2012-01-01

... spinthariscopes), electron tubes, lightning rods, ionization sources, static eliminators, or as designated by the..., or transfer, in accordance with the provisions of paragraphs (b), (c), and (d) of this section, radium-226 contained in the following products manufactured prior to November 30, 2007. (1) Antiquities...

Failure Analysis of Cracked FS-85 Tubing and ASTAR-811C End Caps

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

ME Petrichek

2006-02-09

Failure analyses were performed on cracked FS-85 tubing and ASTAR-811C and caps which had been fabricated as components of biaxial creep specimens meant to support materials testing for the NR Space program. During the failure analyses of cracked FS-85 tubing, it was determined that the failure potentially could be due to two effects: possible copper contamination from the EDM (electro-discharge machined) recast layer and/or an insufficient solution anneal. to prevent similar failures in the future, a more formal analysis should be done after each processing step to ensure the quality of the material before further processing. During machining of themore » ASTAR-811FC rod to form end caps for biaxial creep specimens, linear defects were observed along the center portion of the end caps. These defects were only found in material that was processed from the top portion of the ingot. The linear defects were attributed to a probable residual ingot pipe that was not removed from the ingot. During the subsequent processing of the ingot to rod, the processing temperatures were not high enough to allow self healing of the ingot's residual pipe defect. To prevent this from occurring in the future, it is necessary to ensure that complete removal of the as-melted ingot pipe is verified by suitable non-destructive evaluation (NDE).« less

Distribution of transverse chain fluctuations in harmonically confined semiflexible polymers

NASA Astrophysics Data System (ADS)

Sharma, Rati; Cherayil, Binny J.

2012-05-01

Two different experimental studies of polymer dynamics based on single-molecule fluorescence imaging have recently found evidence of heterogeneities in the widths of the putative tubes that surround filaments of F-actin during their motion in concentrated solution. In one [J. Glaser, D. Chakraborty, K. Kroy, I. Lauter, M. Degawa, N. Kirchesner, B. Hoffmann, R. Merkel, and M. Giesen, Phys. Rev. Lett. 105, 037801 (2010)], 10.1103/PhysRevLett.105.037801, the observations were explained in terms of the statistics of a worm-like chain confined to a potential determined self-consistently by a binary collision approximation, and in the other [B. Wang, J. Guan, S. M. Anthony, S. C. Bae, K. S. Schweizer, and S. Granick, Phys. Rev. Lett. 104, 118301 (2010)], 10.1103/PhysRevLett.104.118301, they were explained in terms of the scaling properties of a random fluid of thin rods. In this paper, we show, using an exact path integral calculation, that the distribution of the length-averaged transverse fluctuations of a harmonically confined weakly bendable rod (one possible realization of a semiflexible chain in a tube), is in good qualitative agreement with the experimental data, although it is qualitatively different in analytic structure from the earlier theoretical predictions. We also show that similar path integral techniques can be used to obtain an exact expression for the time correlation function of fluctuations in the tube cross section.

Nuclear-radiation-actuated valve. [Patent application; for increasing coolant flow to blanket

DOEpatents

Christiansen, D.W.; Schively, D.P.

1982-01-19

The present invention relates to a breeder reactor blanket fuel assembly coolant system valve which increases coolant flow to the blanket fuel assembly to minimize long-term temperature increases caused by fission of fissile fuel created from fertile fuel through operation of the breeder reactor. The valve has a valve first part (such as a valve rod with piston) and a valve second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics. The valve's first part is positioned to receive nuclear radiation from the nuclear reactor's fuel region. The valve's second part is positioned so that its nuclear radiation induced swelling is different from that of the valve's first part. The valve's second part also is positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system.

The Corrosion and Corrosion Fatigue Behavior of Nickel Based Alloy Weld Overlay and Coextruded Claddings

NASA Astrophysics Data System (ADS)

Stockdale, Andrew

The use of low NOx boilers in coal fired power plants has resulted in sulfidizing corrosive conditions within the boilers and a reduction in the service lifetime of the waterwall tubes. As a solution to this problem, Ni-based weld overlays are used to provide the necessary corrosion resistance however; they are susceptible to corrosion fatigue. There are several metallurgical factors which give rise to corrosion fatigue that are associated with the localized melting and solidification of the weld overlay process. Coextruded coatings offer the potential for improved corrosion fatigue resistance since coextrusion is a solid state coating process. The corrosion and corrosion fatigue behavior of alloy 622 weld overlays and coextruded claddings was investigated using a Gleeble thermo-mechanical simulator retrofitted with a retort. The experiments were conducted at a constant temperature of 600°C using a simulated combustion gas of N2-10%CO-5%CO2-0.12%H 2S. An alternating stress profile was used with a minimum tensile stress of 0 MPa and a maximum tensile stress of 300 MPa (ten minute fatigue cycles). The results have demonstrated that the Gleeble can be used to successfully simulate the known corrosion fatigue cracking mechanism of Ni-based weld overlays in service. Multilayer corrosion scales developed on each of the claddings that consisted of inner and outer corrosion layers. The scales formed by the outward diffusion of cations and the inward diffusion of sulfur and oxygen anions. The corrosion fatigue behavior was influenced by the surface finish and the crack interactions. The initiation of a large number of corrosion fatigue cracks was not necessarily detrimental to the corrosion fatigue resistance. Finally, the as-received coextruded cladding exhibited the best corrosion fatigue resistance.

Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

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

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

1991-12-31

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each ofmore » the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.« less

Quartz ball valve

NASA Technical Reports Server (NTRS)

Goetz, C.; Ingle, W. M. (Inventor)

1980-01-01

A ball valve particularly suited for use in the handling of highly corrosive fluids is described. It is characterized by a valve housing formed of communicating segments of quartz tubing, a pair of communicating sockets disposed in coaxial alignment with selected segments of tubing for establishing a pair of inlet ports communicating with a common outlet port, a ball formed of quartz material supported for displacement between the sockets and configured to be received alternately thereby, and a valve actuator including a rod attached to the ball for selectively displacing the ball relative to each of the sockets for controlling fluid flow through the inlet ports.

Study on bubbly flow behavior in natural circulation reactor by thermal-hydraulic simulation tests with SF6-Gas and ethanol liquid

NASA Astrophysics Data System (ADS)

Kondo, Yoshiyuki; Suga, Keishi; Hibi, Koki; Okazaki, Toshihiko; Komeno, Toshihiro; Kunugi, Tomoaki; Serizawa, Akimi; Yoneda, Kimitoshi; Arai, Takahiro

2009-02-01

An advanced experimental technique has been developed to simulate two-phase flow behavior in a light water reactor (LWR). The technique applies three kinds of methods; (1) use of sulfur-hexafluoride (SF6) gas and ethanol (C2H5OH) liquid at atmospheric temperature and a pressure less than 1.0MPa, where the fluid properties are similar to steam-water ones in the LWR, (2) generation of bubble with a sintering tube, which simulates bubble generation on heated surface in the LWR, (3) measurement of detailed bubble distribution data with a bi-optical probe (BOP), (4) and measurement of liquid velocities with the tracer liquid. This experimental technique provides easy visualization of flows by using a large scale experimental apparatus, which gives three-dimensional flows, and measurement of detailed spatial distributions of two-phase flow. With this technique, we have carried out experiments simulating two-phase flow behavior in a single-channel geometry, a multi-rod-bundle one, and a horizontal-tube-bundle one on a typical natural circulation reactor system. Those experiments have clarified a) a flow regime map in a rod bundle on the transient region between bubbly and churn flow, b) three-dimensional flow behaviour in rod-bundles where inter-subassembly cross-flow occurs, c) bubble-separation behavior with consideration of reactor internal structures. The data have given analysis models for the natural circulation reactor design with good extrapolation.

Data summary report for fission product release test VI-6

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

Osborne, M.F.; Lorenz, R.A.; Travis, J.R.

Test VI-6 was the sixth test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium. The fuel had experienced a burnup of {approximately}42 MWd/kg, with inert gas release during irradiation of {approximately}2%. The fuel specimen was heated in an induction furnace at 2300 K for 60 min, initially in hydrogen, then in a steam atmosphere. The released fission products were collected in three sequentially operated collection trains designed to facilitate sampling and analysis. The fission product inventories in the fuel were measured directlymore » by gamma-ray spectrometry, where possible, and were calculated by ORIGEN2. Integral releases were 75% for {sup 85}Kr, 67% for {sup 129}I, 64% for {sup 125}Sb, 80% for both {sup 134}Cs and {sup 137}Cs, 14% for {sup 154}Eu, 63% for Te, 32% for Ba, 13% for Mo, and 5.8% for Sr. Of the totals released from the fuel, 43% of the Cs, 32% of the Sb, and 98% of the Eu were deposited in the outlet end of the furnace. During the heatup in hydrogen, the Zircaloy cladding melted, ran down, and reacted with some of the UO{sub 2} and fission products, especially Te and Sb. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.57 g, almost equally divided between thermal gradient tubes and filters. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL Diffusion Model.« less

Stimulated Raman scattering in AsSe2-As2S5 microstructured optical fiber

NASA Astrophysics Data System (ADS)

Gao, Weiqing; Ni, Chenquan; Xu, Qiang; Li, Xue; Chen, Xiangcai; Chen, Li; Wen, Zhenqiang; Cheng, Tonglei; Xue, Xiaojie; Suzuki, Takenobu; Ohishi, Yasutake

2017-02-01

We demonstrate the effects of stimulated Raman scattering (SRS) in the all-solid-core chalcogenide microstructured optical fibers (MOFs) with AsSe2 core and As2S5 cladding, which are fabricated by the rod-in-tube drawing technique. The core diameters of the MOFs are 6.3 (Fiber I), 3.0 (Fiber II), 2.6 (Fiber III) and 2.2 (Fiber IV) μm, respectively. The chromatic dispersion of the fundamental mode in Fibers I-IV is simulated by the full-vectorial mode solver technique. The first-order Stokes wave is investigated in the fibers with different core diameters pumped by the picosecond pulses at 1958 nm. In Fiber I, no obvious Raman peak is observed with the pump power increasing, because the effective nonlinearity is not high. In Fiber II, a Raman Stokes peak at 2065 nm begins to emerge at the pump power of 110 mW. The conversion efficiency is as weak as -36.6 dB at 150 mW pumping. In Fiber III, the first-order Raman peak at 2060 nm begins to emerge at 40 mW pumping. The conversion efficiency is -15.0 dB, which is 21.6 dB higher than that in Fiber II. In Fiber IV, the Stokes peak at 2070 nm begins to appear at 56 mW pumping. The maximum conversion efficiency of the first-order Stokes wave is obtained in the MOF with the core diameter of 2.6 μm. The evolution of the first-order Stokes wave with pump power and fiber length is investigated. This is the first demonstration of Raman effects in the AsSe2-As2S5 MOF, to the best of our knowledge.

Recording of Electric Signal Passing Through a Pylon in Direct Skeletal Attachment of Leg Prostheses with Neuromuscular Control

PubMed Central

Pitkin, M.; Cassidy, C.; Muppavarapu, R.; Edell, David

2012-01-01

Direct recordings were made of electrical signals emanating from the muscles in a rabbit’s residuum. The signals were transmitted via wires attached on one end to the muscles, and on the other to an external recording system. The cable was held in a titanium tube inside a pylon that had been transcutaneously implanted into the residuum’s bone. The tube was surrounded by porous titanium cladding to enhance its bond with the bone and with the skin of the residuum. This study was the first known attempt to merge the technology of direct skeletal attachment of limb prostheses with the technology of neuromuscular control of prostheses, providing a safe and reliable passage of the electrical signal from the muscles inside the residuum to the outside recording system. PMID:22345523

Diffusion and Interface Effects during Preparation of All-Solid Microstructured Fibers

PubMed Central

Jens, Kobelke; Jörg, Bierlich; Katrin, Wondraczek; Claudia, Aichele; Zhiwen, Pan; Sonja, Unger; Kay, Schuster; Hartmut, Bartelt

2014-01-01

All-solid microstructured optical fibers (MOF) allow the realization of very flexible optical waveguide designs. They are prepared by stacking of doped silica rods or canes in complex arrangements. Typical dopants in silica matrices are germanium and phosphorus to increase the refractive index (RI), or boron and fluorine to decrease the RI. However, the direct interface contact of stacking elements often causes interrelated chemical reactions or evaporation during thermal processing. The obtained fiber structures after the final drawing step thus tend to deviate from the targeted structure risking degrading their favored optical functionality. Dopant profiles and design parameters (e.g., the RI homogeneity of the cladding) are controlled by the combination of diffusion and equilibrium conditions of evaporation reactions. We show simulation results of diffusion and thermal dissociation in germanium and fluorine doped silica rod arrangements according to the monitored geometrical disturbances in stretched canes or drawn fibers. The paper indicates geometrical limits of dopant structures in sub-µm-level depending on the dopant concentration and the thermal conditions during the drawing process. The presented results thus enable an optimized planning of the preform parameters avoiding unwanted alterations in dopant concentration profiles or in design parameters encountered during the drawing process. PMID:28788219

Diffusion and Interface Effects during Preparation of All-Solid Microstructured Fibers.

PubMed

Jens, Kobelke; Jörg, Bierlich; Katrin, Wondraczek; Claudia, Aichele; Zhiwen, Pan; Sonja, Unger; Kay, Schuster; Hartmut, Bartelt

2014-09-25

All-solid microstructured optical fibers (MOF) allow the realization of very flexible optical waveguide designs. They are prepared by stacking of doped silica rods or canes in complex arrangements. Typical dopants in silica matrices are germanium and phosphorus to increase the refractive index (RI), or boron and fluorine to decrease the RI. However, the direct interface contact of stacking elements often causes interrelated chemical reactions or evaporation during thermal processing. The obtained fiber structures after the final drawing step thus tend to deviate from the targeted structure risking degrading their favored optical functionality. Dopant profiles and design parameters (e.g., the RI homogeneity of the cladding) are controlled by the combination of diffusion and equilibrium conditions of evaporation reactions. We show simulation results of diffusion and thermal dissociation in germanium and fluorine doped silica rod arrangements according to the monitored geometrical disturbances in stretched canes or drawn fibers. The paper indicates geometrical limits of dopant structures in sub-µm-level depending on the dopant concentration and the thermal conditions during the drawing process. The presented results thus enable an optimized planning of the preform parameters avoiding unwanted alterations in dopant concentration profiles or in design parameters encountered during the drawing process.

Polymer Nanocomposites with Prescribed Morphology: Going Beyond Nanoparticle-Filled Polymers (Preprint)

DTIC Science & Technology

2006-10-01

evidence for these parallels through the depression of glass transition temperature in silica-filled polystyrene.31 For the final material performance...has been reported for molecular nanowires, tubes, ribbons, and rods (Li2Mo6- Se6, imogolite, Nb2PS10, V2O5 , boehmite (γ-AlOOH), aka- ganeite (â-FeOOH

AN INSTRUMENT TO MEASURE THE ELECTRICAL CONDUCTIVITY OF ARC PLASMA JETS

DTIC Science & Technology

The instrument was calibrated by moving aluminum or graphite rods through the transducer. By using thin-wall, stainless steel tubing, the influence ... function for the transducer was also obtained. Tests were run on two different arc plasma jet facilities. Values of s, u ranged from 0.02 to 9 megamhos per second. (Author)