Compact Hybrid Laser Rod and Laser System

NASA Technical Reports Server (NTRS)

Pierrottet, Diego F. (Inventor); Busch, George E. (Inventor); Amzajerdian, Farzin (Inventor)

2017-01-01

A hybrid fiber rod includes a fiber core and inner and outer cladding layers. The core is doped with an active element. The inner cladding layer surrounds the core, and has a refractive index substantially equal to that of the core. The outer cladding layer surrounds the inner cladding layer, and has a refractive index less than that of the core and inner cladding layer. The core length is about 30 to 2000 times the core diameter. A hybrid fiber rod laser system includes an oscillator laser, modulating device, the rod, and pump laser diode(s) energizing the rod from opposite ends. The rod acts as a waveguide for pump radiation but allows for free-space propagation of laser radiation. The rod may be used in a laser resonator. The core length is less than about twice the Rayleigh range. Degradation from single-mode to multi-mode beam propagation is thus avoided.

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

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.

Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

NASA Astrophysics Data System (ADS)

Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

2018-02-01

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

The behavior of breached boiling water reactor fuel rods on long-term exposure to air and argon at 598 K

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

Kohli, R.; Gilbert, E.R.; Johnson, A.B.

1985-05-01

Two irradiated boiling water reactor fuel rods with breached cladding were exposed to argon and to air at 598 K for 7.56 Ms (2100 h). These tests were conducted to determine fuel swelling and cladding crack propagation under conditions that promote UO/sub 2/ fuel oxidation and to observe the behavior of water-logged breached fuel in an inert gas environment. The two rods were selected for testing after extensive hot cell examination had shown the cladding of both rods to be breached with several centimetres of open cracks; the cracks were characterized in detail before the test. As part of themore » experiment, the amount of the readily removable water contained in the fuel rods was determined. To oxidize the fuel to a significant level ( about10%), the air in the annealine capsule was replenished approximately daily. The depletion of oxygen available in the air capsule due to fuel oxidation occurred in about0.036 Ms (10 h). At the end of the test period, about6% of the fuel is estimated to have oxidized. Posttest examination of the rods showed that cladding degradation resulted from swelling due to oxidation of the fuel in the air environment. The cladding degradation was localized and fuel oxidation did not measurably extend beyond the cladding breach. No cladding degradation was measurable in the breached fuel rod tested in argon.« less

Review of CTF s Fuel Rod Modeling Using FRAPCON-4.0 s Centerline Temperature Predictions

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

Toptan, Aysenur; Salko, Robert K; Avramova, Maria

Coolant Boiling in Rod Arrays Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF s fuel rod modeling to address shortcomings in CTF s temperature predictions. CTF is compared to FRAPCON which is U.S. NRC s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes inmore » fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF s dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF s dynamic gap conductance model. Improving the CTF s dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.« less

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

DOE PAGES

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

2016-10-01

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

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

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

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

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

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.

Fuel cladding behavior under rapid loading conditions

NASA Astrophysics Data System (ADS)

Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

2016-02-01

A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

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

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

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.

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

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

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

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.

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

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

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.

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

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.

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

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

Patra, Anirban; Tomé, Carlos N.

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

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

DOE PAGES

Patra, Anirban; Tomé, Carlos N.

2017-03-06

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

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

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.

Irradiation effects on thermal properties of LWR hydride fuel

NASA Astrophysics Data System (ADS)

Terrani, Kurt; Balooch, Mehdi; Carpenter, David; Kohse, Gordon; Keiser, Dennis; Meyer, Mitchell; Olander, Donald

2017-04-01

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

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.

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

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.

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

TEST SYSTEM FOR EVALUATING SPENT NUCLEAR FUEL BENDING STIFFNESS AND VIBRATION INTEGRITY

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

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

2013-01-01

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements specified by federal regulations. For normal conditions of transport, vibration loads incident to transport must be considered. This is particularly relevant for high-burnup fuel (>45 GWd/MTU). As the burnup of the fuel increases, a number of changes occur that may affect the performance of the fuel and cladding in storage and during transportation. The mechanical properties of high-burnup de-fueled cladding have been previously studied by subjecting defueled cladding tubes to longitudinal (axial) tensile tests, ring-stretch tests, ring-compression tests, and biaxial tube burst tests. The objective of this study ismore » to investigate the mechanical properties and behavior of both the cladding and the fuel in it under vibration/cyclic loads similar to the sustained vibration loads experienced during normal transport. The vibration loads to SNF rods during transportation can be characterized by dynamic, cyclic, bending loads. The transient vibration signals in a specified transport environment can be analyzed, and frequency, amplitude and phase components can be identified. The methodology being implemented is a novel approach to study the vibration integrity of actual SNF rod segments through testing and evaluating the fatigue performance of SNF rods at defined frequencies. Oak Ridge National Laboratory (ORNL) has developed a bending fatigue system to evaluate the response of the SNF rods to vibration loads. A three-point deflection measurement technique using linear variable differential transformers is used to characterize the bending rod curvature, and electromagnetic force linear motors are used as the driving system for mechanical loading. ORNL plans to use the test system in a hot cell for SNF vibration testing on high burnup, irradiated fuel to evaluate the pellet-clad interaction and bonding on the effective lifetime of fuel-clad structure bending fatigue performance. Technical challenges include pure bending implementation, remote installation and detachment of the SNF test specimen, test specimen deformation measurement, and identification of a driving system suitable for use in a hot cell. Surrogate test specimens have been used to calibrate the test setup and conduct systematic cyclic tests. The calibration and systematic cyclic tests have been used to identify test protocol issues prior to implementation in the hot cell. In addition, cyclic hardening in unidirectional bending and softening in reverse bending were observed in the surrogate test specimens. The interface bonding between the surrogate clad and pellets was found to impact the bending response of the surrogate rods; confirming this behavior in the actual spent fuel segments will be an important aspect of the hot cell test implementation,« less

Cladding burst behavior of Fe-based alloys under LOCA

DOE PAGES

Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; ...

2015-12-17

Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. Themore » most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.« less

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.

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

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

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.

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.

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.

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

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

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

Modified rod-in-tube for high-NA tellurite glass fiber fabrication: materials and technologies.

PubMed

Chen, Qiuling; Wang, Hui; Wang, Qingwei; Chen, Qiuping; Hao, Yinlei

2015-02-01

In this paper, we report the whole fabrication process for high-numerical aperture (NA) tellurite glass fibers from material preparation to preform fabrication, and eventually, fiber drawing. A tellurite-based high-NA (0.9) magneto-optical glass fiber was drawn successfully and characterized. First, matchable core and cladding glasses were fabricated and matched in terms of physical properties. Second, a uniform bubble-free preform was fabricated by means of a modified rod-in-tube technique. Finally, the fiber drawing process was studied and optimized. The high-NA fibers (∅(core), 40-50 μm and ∅(cladding), 120-130 μm) so obtained were characterized for their geometrical and optical properties.

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

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

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

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.

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 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.

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

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

Charpenel, Jean; Lemoine, Francette; Sato, Ikken

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

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

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.

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

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

First overpower tests of metallic IFR [Integral Fast Reactor] fuel in TREAT [Transient Reactor Test Facility]: Data and analysis from tests M5, M6, and M7

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

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

1989-12-01

Results and analyses of margin to cladding failure and pre-failure axial expansion of metallic fuel are reported for TREAT in-pile transient overpower tests M5--M7. These are the first such tests on reference binary and ternary alloy fuel of the Integral Fast Reactor (IFR) concept with burnup ranging from 1 to 10 at. %. In all cases, test fuel was subjected to an exponential power rise on an 8 s period until either incipient or actual cladding failure was achieved. Objectives, designs and methods are described with emphasis on developments unique to metal fuel safety testing. The resulting database for claddingmore » failure threshold and prefailure fuel expansion is presented. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models of cladding failures and pre-failure axial expansions are described and compared with experimental results. Reported results include: temperature, flow, and pressure data from test instrumentation; fuel motion diagnostic data principally from the fast neutron hodoscope; and test remains described from both destructive and non-destructive post-test examination. 24 refs., 144 figs., 17 tabs.« less

The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

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

Wang, Jy-An John; Jiang, Hao

2016-01-12

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (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 using 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 themore » simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.« less

Fabrication of light water reactor tritium targets

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

Pilger, J.P.

1991-11-01

The mission of the Fabrication Development Task of the Tritium Target Development Project is: to produce a documented technology basis, including specifications and procedures for target rod fabrication; to demonstrate that light water tritium targets can be manufactured at a rate consistent with tritium production requirements; and to develop quality control methods to evaluate target rod components and assemblies, and establish correlations between evaluated characteristics and target rod performance. Many of the target rod components: cladding tubes, end caps, plenum springs, etc., have similar counterparts in LWR fuel rods. High production rate manufacture and inspection of these components has beenmore » adequately demonstrated by nuclear fuel rod manufacturers. This summary describes the more non-conventional manufacturing processes and inspection techniques developed to fabricate target rod components whose manufacturability at required production rates had not been previously demonstrated.« 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.

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.

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

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

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

1979-03-22

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

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

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

On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

PubMed

Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

2016-08-01

For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. Copyright © 2016 Elsevier Ltd. All rights reserved.

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.

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).

Microstructure and Mechanical Properties of Laser Clad and Post-cladding Tempered AISI H13 Tool Steel

NASA Astrophysics Data System (ADS)

Telasang, Gururaj; Dutta Majumdar, Jyotsna; Wasekar, Nitin; Padmanabham, G.; Manna, Indranil

2015-05-01

This study reports a detailed investigation of the microstructure and mechanical properties (wear resistance and tensile strength) of hardened and tempered AISI H13 tool steel substrate following laser cladding with AISI H13 tool steel powder in as-clad and after post-cladding conventional bulk isothermal tempering [at 823 K (550 °C) for 2 hours] heat treatment. Laser cladding was carried out on AISI H13 tool steel substrate using a 6 kW continuous wave diode laser coupled with fiber delivering an energy density of 133 J/mm2 and equipped with a co-axial powder feeding nozzle capable of feeding powder at the rate of 13.3 × 10-3 g/mm2. Laser clad zone comprises martensite, retained austenite, and carbides, and measures an average hardness of 600 to 650 VHN. Subsequent isothermal tempering converted the microstructure into one with tempered martensite and uniform dispersion of carbides with a hardness of 550 to 650 VHN. Interestingly, laser cladding introduced residual compressive stress of 670 ± 15 MPa, which reduces to 580 ± 20 MPa following isothermal tempering. Micro-tensile testing with specimens machined from the clad zone across or transverse to cladding direction showed high strength but failure in brittle mode. On the other hand, similar testing with samples sectioned from the clad zone parallel or longitudinal to the direction of laser cladding prior to and after post-cladding tempering recorded lower strength but ductile failure with 4.7 and 8 pct elongation, respectively. Wear resistance of the laser surface clad and post-cladding tempered samples (evaluated by fretting wear testing) registered superior performance as compared to that of conventional hardened and tempered AISI H13 tool steel.

Fuel rod with annular nuclear fuel pellets having same U-235 enrichment and different annulus sizes for graduated enrichment loading

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

Mildrum, C.M.

1987-08-18

A fuel rod is described for a nuclear reactor fuel assembly, comprising: (a) a hollow cladding tube; (b) a pair of end plugs connected to and sealing the cladding tube at opposite ends thereof; (c) a plurality of fuel pellets contained on the tube and being composed of fissile material having a single enrichment the value of which is at the level of the maximum enrichment loading of the rod, the pellets having provided in a stack having one end disposed adjacent to one of the end plugs and an opposite end disposed remote from the other of the endmore » plugs; and (d) a plenum spring disposed in the tube between the other end plug and the opposite end of the pellet stack for retaining the pellets in a stack form; (e) at least some of the fuel pellets having an annular configuration and at least other of the fuel pellets having a solid configuration; (f) each of some of the annular fuel pellets having an annulus of a first size; (e) each of other of the annual fuel pellets having an annulus of a second size different from the first size, whereby graduation of axial enrichment loading is provided between the annual fuel pellets of the fuel rod.« 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

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

FPIN2 posttest analysis of cylindrical canisters in SLSF Experiment P4

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

Hughes, T H; Kramer, J M

Results demonstrate that the clad deformation is dominated by the expansion of the fuel when it melts. In our analysis we moved the end space volume and some of the fuel-clad radial gap volume to an artificial central hole. This approximation may affect the details in the early parts of the transient, but clearly did not affect the major cladding deformation. It is also clear that the accuracy of the value of the fuel expansion upon melting is significant as is the dimensional accuracy of the fuel and canisters. The major conclusions from the FPIN2 posttest analysis of the cylindricalmore » canisters in SLSF Experiment P4 are: The maximum melt fractions in the two canisters were about 75%. Both canisters experienced about the same diametral strains of 12% prior to failure. These strains were almost entirely due to the additional volume that must be created inside the canisters to accommodate the expansion of fuel on melting. The mode of cladding failure was plastic instability by necking of the canister walls. The failure time of the 20% CW canister and the nonmechanical failure of the 10% CW canister are consistent with the FPIN2 calculations using the plastic instability failure criteria.« less

Development of Safety Analysis Code System of Beam Transport and Core for Accelerator Driven System

NASA Astrophysics Data System (ADS)

Aizawa, Naoto; Iwasaki, Tomohiko

2014-06-01

Safety analysis code system of beam transport and core for accelerator driven system (ADS) is developed for the analyses of beam transients such as the change of the shape and position of incident beam. The code system consists of the beam transport analysis part and the core analysis part. TRACE 3-D is employed in the beam transport analysis part, and the shape and incident position of beam at the target are calculated. In the core analysis part, the neutronics, thermo-hydraulics and cladding failure analyses are performed by the use of ADS dynamic calculation code ADSE on the basis of the external source database calculated by PHITS and the cross section database calculated by SRAC, and the programs of the cladding failure analysis for thermoelastic and creep. By the use of the code system, beam transient analyses are performed for the ADS proposed by Japan Atomic Energy Agency. As a result, the rapid increase of the cladding temperature happens and the plastic deformation is caused in several seconds. In addition, the cladding is evaluated to be failed by creep within a hundred seconds. These results have shown that the beam transients have caused a cladding failure.

Improving the Thermal Shock Resistance of Thermal Barrier Coatings Through Formation of an In Situ YSZ/Al2O3 Composite via Laser Cladding

NASA Astrophysics Data System (ADS)

Soleimanipour, Zohre; Baghshahi, Saeid; Shoja-razavi, Reza

2017-04-01

In the present study, laser cladding of alumina on the top surface of YSZ thermal barrier coatings (TBC) was conducted via Nd:YAG pulsed laser. The thermal shock behavior of the TBC before and after laser cladding was modified by heating at 1000 °C for 15 min and quenching in cold water. Phase analysis, microstructural evaluation and elemental analysis were performed using x-ray diffractometry, scanning electron microscopy (SEM), and energy-dispersive spectroscopy. The results of thermal shock tests indicated that the failure in the conventional YSZ (not laser clad) and the laser clad coatings happened after 200 and 270 cycles, respectively. The SEM images of the samples showed that delamination and spallation occurred in both coatings as the main mechanism of failure. Formation of TGO was also observed in the fractured cross section of the samples, which is also a main reason for degradation. Thermal shock resistance in the laser clad coatings improved about 35% after cladding. The improvement is due to the presence of continuous network cracks perpendicular to the surface in the clad layer and also the thermal stability and high melting point of alumina in Al2O3/ZrO2 composite.

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.

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.

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.

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

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

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

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

Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

NASA Astrophysics Data System (ADS)

Li, Bo-Shiuan

Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission gas release. An optimum design is sought considering both thermal and mechanical models of this ceramic cladding with UO2 and advanced high density fuels.

Experience reveals ways to minimize failures in rod-pumped wells

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

Patterson, J.C.; Bucaram, S.M.; Curfew, J.V.

From the experience gained over the past 25 years, ARCO Oil and Gas Co. has developed recommendations to reduce equipment failure in sucker-rod pumping installations. These recommendations include equipment selection and design, operating procedures, and chemical treatment. Equipment failure and its attendant costs are extremely important in today's petroleum industry. Because rod pumping is the predominant means of artificial lift, minimizing equipment failure in rod pumped wells can have a significant impact on profitability. This compilation of recommendations comes from field locations throughout the US and other countries. The goal is to address and solve problems on a well-by-well basis.

Posttest data analysis of FIST experimental TRAC-BD1/MOD1 power transient experiment

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

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting in only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena: (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« 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...

Retrieval and clinical analysis of distraction-based dual growing rod constructs for early-onset scoliosis.

PubMed

Hill, Genevieve; Nagaraja, Srinidhi; Akbarnia, Behrooz A; Pawelek, Jeff; Sponseller, Paul; Sturm, Peter; Emans, John; Bonangelino, Pablo; Cockrum, Joshua; Kane, William; Dreher, Maureen

2017-10-01

Growing rod constructs are an important contribution for treating patients with early-onset scoliosis. These devices experience high failure rates, including rod fractures. The objective of this study was to identify the failure mechanism of retrieved growing rods, and to identify differences between patients with failed and intact constructs. Growing rod patients who had implant removal and were previously enrolled in a multicenter registry were eligible for this study. Forty dual-rod constructs were retrieved from 36 patients across four centers, and 34 of those constructs met the inclusion criteria. Eighteen constructs failed due to rod fracture. Sixteen intact constructs were removed due to final fusion (n=7), implant exchange (n=5), infection (n=2), or implant prominence (n=2). Analyses of clinical registry data, radiographs, and retrievals were the outcome measures. Retrievals were analyzed with microscopic imaging (optical and scanning electron microscopy) for areas of mechanical failure, damage, and corrosion. Failure analyses were conducted on the fracture surfaces to identify failure mechanism(s). Statistical analyses were performed to determine significant differences between the failed and intact groups. The failed rods fractured due to bending fatigue under flexion motion. Construct configuration and loading dictate high bending stresses at three distinct locations along the construct: (1) mid-construct, (2) adjacent to the tandem connector, or (3) adjacent to the distal anchor foundation. In addition, high torques used to insert set screws may create an initiation point for fatigue. Syndromic scoliosis, prior rod fractures, increase in patient weight, and rigid constructs consisting of tandem connectors and multiple crosslinks were associated with failure. This is the first study to examine retrieved, failed growing rod implants across multiple centers. Our analysis found that rod fractures are due to bending fatigue, and that stress concentrations play an important role in rod fractures. Recommendations are made on surgical techniques, such as the use of torque-limiting wrenches or not exceeding the prescribed torques. Additional recommendations include frequent rod replacement in select patients during scheduled surgeries. Published by Elsevier Inc.

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.

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.

Advanced Metalworking Solutions for Naval Systems That Go in Harm’s Way

DTIC Science & Technology

2008-01-01

and stress corrosion cracking failures; unique processes using compacted titanium powders that are subsequently flowformed into piping, thereby...damping characteristics, adhesion strengths in peel and shock, toxicity, flame retardancy and others. Alternate cladding techniques are being...which produces a high-quality clad overlay but at a low deposition rate relative to other cladding processes. NMC and the project team will

Failure and penetration response of borosilicate glass during short-rod impact

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

Anderson, C. E. Jr.; Orphal, D. L.; Behner, Th.

2007-12-12

The failure characterization of brittle materials like glass is of fundamental importance in describing the penetration resistance against projectiles. A critical question is whether this failure front remains 'steady' after the driving stress is removed. A test series with short gold rods (D = 1 mm, L/D{approx_equal}5-11) impacting borosilicate glass at {approx}1 to 2 km/s was carried out to investigate this question. The reverse ballistic method was used for the experiments, and the impact and penetration process was observed simultaneously with five flash X-rays and a 16-frame high-speed optical camera. Very high measurement accuracy was established to ensure reliable results.more » Results show that the failure front induced by rod impact and penetration does arrest (ceases to propagate) after the rod is totally eroded inside the glass. The impact of a second rod after a short time delay reinitiates the failure front at about the same speed.« less

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Nuclear fuel elements having a composite cladding

DOEpatents

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

1983-09-20

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

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

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

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

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.

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

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.

Ultrasonic Structural Health Monitoring to Assess the Integrity of Spinal Growing Rods In Vitro.

PubMed

Oetgen, Matthew E; Goodley, Addison; Yoo, Byungseok; Pines, Darryll J; Hsieh, Adam H

2016-01-01

Rod fracture is a common complication of growing rods and can result in loss of correction, patient discomfort, and unplanned revision surgery. The ability to quantitate rod integrity at each lengthening would be advantageous to avoid this complication. We investigate the feasibility of applying structural health monitoring to evaluate the integrity of growing rods in vitro. Single-rod titanium 4.5-mm growing rod constructs (n = 9), one screw proximally and one distally connected by in-line connectors, were assembled with pedicle screws fixed in polyethylene blocks. Proximal and distal ends were loaded and constructs subjected to cyclic axial compression (0-100 N at 1 Hz), with incrementally increasing maximum compressive loads of 10 N every 9k cycles until failure. Four piezoceramic transducers (PZTs) were mounted along the length the constructs to interrogate the integrity of the rods with an ultrasonic, guided lamb wave approach. Every 9k cycles, an 80 V excitatory voltage was applied to a PZT to generate high-frequency vibrations, which, after propagating through the construct, was detected by the remaining PZTs. Amplitude differences between pre- and postload waveform signals were calculated until rod failure. Average construct lifetime was 88,991 ± 13,398 cycles. All constructs failed due to rod fracture within 21 mm (mean = 15 ± 4.5 mm) of a screw or connector. Amplitude differences between pre- and postload increased in a stepwise fashion as constructs were cycled. Compared to baseline, we found a 1.8 ± 0.6-fold increase in amplitude 18k cycles before failure, a 2.2 ± 1.0-fold increase in amplitude 9k cycles before failure, and a 2.75 ± 1.5-fold increase in amplitude immediately before rod fracture. We describe a potential method for assessing the structural integrity of growing rods using ultrasonic structural health monitoring. These preliminary data demonstrate the ability of periodic rod assessment to detect structural changes in cycled growing rods, which appear to correspond to subclinical rod fatigue before rod fracture. Copyright © 2016 Scoliosis Research Society. Published by Elsevier Inc. All rights reserved.

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.

Study of the Optimum Zone of the Independent Variables of an ORGEL Reactor Connected to a 250-MWeb Power Plant. Self Supporting Fuel Elements Made of UC, with Sap Cladding with Four Fuel Rods and Individual Pressure Tubes; STUDIE DER OPTIMALEN ZONE DER UNABHANGIGEN PARAMETER EINES ORGEL- REAKTORS IN EINEM 250-MWe-KRAFTWERK. SELBST-TRAGENDES BRENNELEMENT AUS UC, SAP-UMHUL-LUNG MIT 4 BRENNSTOFFSTABEN UND INDIVIDUELLEN DRUCKROHREN

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

LaFontaine, F.; Tauch, P.

The optimum range of the independent variables of and ORGEL reactor connected to a 250-Mw power plant (4 fuel rods of UC with individual pressure tubes), as well as the geometry of the reactor core and the operation of the plant, is described. (auth)

Posttest data analysis and assessment of TRAC-BD1/MOD1 with data from a Full Integral Simulation Test (FIST) power transient experiment

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

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting on only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena; (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

Ho3+-doped AlF3-TeO2-based glass fibers for 2.1 µm laser applications

NASA Astrophysics Data System (ADS)

Wang, S. B.; Jia, Z. X.; Yao, C. F.; Ohishi, Y.; Qin, G. S.; Qin, W. P.

2017-05-01

Ho3+-doped AlF3-TeO2-based glass fibers based on AlF3-BaF2-CaF2-YF3-SrF2-MgF2-TeO2 glasses are fabricated by using a rod-in-tube method. The glass rod including a core and a thick cladding layer is prepared by using a suction method, where the thick cladding layer is used to protect the core from the effect of surface crystallization during the fiber drawing. By inserting the glass rod into a glass tube, the glass fibers with relatively low loss (~2.3 dB m-1 @ 1560 nm) are prepared. By using a 38 cm long Ho3+-doped AlF3-TeO2-based glass fiber as the gain medium and a 1965 nm fiber laser as the pump source, 2065 nm lasing is obtained for a threshold pump power of ~220 mW. With further increasing the pump power to ~325 mW, the unsaturated output power of the 2065 nm laser is about 82 mW and the corresponding slope efficiency is up to 68.8%. The effects of the gain fiber length on the lasing threshold, the slope efficiency, and the operating wavelength are also investigated. Our experimental results show that Ho3+-doped AlF3-TeO2-based glass fibers are promising gain media for 2.1 µm laser applications.

Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

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

Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.

The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approachmore » to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)« less

A two-scale Weibull approach to the failure of porous ceramic structures made by robocasting: possibilities and limits

PubMed Central

Genet, Martin; Houmard, Manuel; Eslava, Salvador; Saiz, Eduardo; Tomsia, Antoni P.

2012-01-01

This paper introduces our approach to modeling the mechanical behavior of cellular ceramics, through the example of calcium phosphate scaffolds made by robocasting for bone-tissue engineering. The Weibull theory is used to deal with the scaffolds’ constitutive rods statistical failure, and the Sanchez-Palencia theory of periodic homogenization is used to link the rod- and scaffold-scales. Uniaxial compression of scaffolds and three-point bending of rods were performed to calibrate and validate the model. If calibration based on rod-scale data leads to over-conservative predictions of scaffold’s properties (as rods’ successive failures are not taken into account), we show that, for a given rod diameter, calibration based on scaffold-scale data leads to very satisfactory predictions for a wide range of rod spacing, i.e. of scaffold porosity, as well as for different loading conditions. This work establishes the proposed model as a reliable tool for understanding and optimizing cellular ceramics’ mechanical properties. PMID:23439936

Experimental and numerical investigation on cladding of corrosion-erosion resistant materials by a high power direct diode laser

NASA Astrophysics Data System (ADS)

Farahmand, Parisa

In oil and gas industry, soil particles, crude oil, natural gas, particle-laden liquids, and seawater can carry various highly aggressive elements, which accelerate the material degradation of component surfaces by combination of slurry erosion, corrosion, and wear mechanisms. This material degradation results into the loss of mechanical properties such as strength, ductility, and impact strength; leading to detachment, delamination, cracking, and ultimately premature failure of components. Since the failure of high valued equipment needs considerable cost and time to be repaired or replaced, minimizing the tribological failure of equipment under aggressive environment has been gaining increased interest. It is widely recognized that effective management of degradation mechanisms will contribute towards the optimization of maintenance, monitoring, and inspection costs. The hardfacing techniques have been widely used to enhance the resistance of surfaces against degradation mechanisms. Applying a surface coating improves wear and corrosion resistance and ensures reliability and long-term performance of coated parts. A protective layer or barrier on the components avoids the direct mechanical and chemical contacts of tool surfaces with process media and will reduce the material loss and ultimately its failure. Laser cladding as an advanced hardfacing technique has been widely used for industrial applications in order to develop a protective coating with desired material properties. During the laser cladding, coating material is fused into the base material by means of a laser beam in order to rebuild a damaged part's surface or to enhance its surface function. In the hardfacing techniques such as atmospheric plasma spraying (APS), high velocity oxygen-fuel (HVOF), and laser cladding, mixing of coating materials with underneath surface has to be minimized in order to utilize the properties of the coating material most effectively. In this regard, laser cladding offers advantages due to creating coating layers with superior properties in terms of purity, homogeneity, low dilution, hardness, bonding, and microstructure. In the development of modern materials for hardfacing applications, the functionality is often improved by combining materials with different properties into composites. Metal Matrix Composite (MMC) coating is a composite material with two constituent parts, i.e., matrix and the reinforcement. This class of composites are addressing improved mechanical properties such as stiffness, strength, toughness, and tribological and chemical resistance. Fabrication of MMCs is to achieve a combination of properties not achievable by any of the materials acting alone. MMCs have attracted significant attention for decades due to their combination of wear-resistivity, corrosion-resistivity, thermal, electrical and magnetic properties. Presently, there is a strong emphasis on the development of advanced functional coatings for corrosion, erosion, and wear protection for different industrial applications. In this research, a laser cladding system equipped with a high power direct diode laser associated with gas driven metal powder delivery system was used to develop advanced MMC coatings. The high power direct diode laser used in this study offers wider beam spot, shorter wavelength and uniform power distribution. These properties make the cladding set-up ideal for coating due to fewer cladding tracks, lower operation cost, higher laser absorption, and improved coating qualities. In order to prevent crack propagation, porosity, and uniform dispersion of carbides in MMC coating, cladding procedure was assisted by an induction heater as a second heat source. The developed defect free MMC coatings were combined with nano-size particles of WC, rare earth (RE) element (La2O3), and Mo as a refractory metal to enhance mechanical properties, chemical composition, and subsequently improve the tribological performance of the coatings. The resistance of developed MMC coatings were examined under highly accelerated slurry erosion, corrosion, and wear as the most frequently encountered failure modes of mechanical components. The microstructure, mechanical properties, and the level of induced residual stress on the coating after cladding procedure are closely related to cladding process variables. Study about the effect of processing parameters on clad quality and experienced thermal history and thermally-induced stress evolution requires both theoretical and experimental understanding of the associated physical phenomena. Numerical modeling offers a cost-efficient way to better understand the related complex physics in laser cladding process. It helps to reveal the effects and significance of each processing parameters on the desired characteristics of clad parts. Successful numerical simulation can provide unique insight into complex laser cladding process, efficiently calculate the complex procedure, and help to obtain coating parts with quality integrity. Therefore, current study develops a three-dimensional (3D) transient and uncoupled thermo-elastic-plastic model to study thermal history, molten pool evolution, thermally induced residual stress, and the effect of utilizing an induction heater as a second heat source on the mechanical properties and microstructural properties of final cladded coating.

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.

76 FR 33176 - Airworthiness Directives; Airbus Model A300 B4-103, B4-203, and B4-2C Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-08

...: One operator reported a failure of the MLG [main landing gear] retraction actuator sliding rod. This...: One operator reported a failure of the MLG [main landing gear] retraction actuator sliding rod. This... inspections of the retraction actuator sliding rod as installed on A300, A300-600 and A300-600ST aeroplanes...

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.

Design solutions for the solar cell interconnect fatigue fracture problem

NASA Technical Reports Server (NTRS)

Mon, G. R.; Ross, R. G., Jr.

1982-01-01

Mechanical fatigue of solar cell interconnects is a major failure mechanism in photovoltaic arrays. A comprehensive approach to the reliability design of interconnects, together with extensive design data for the fatigue properties of copper interconnects, has been published. This paper extends the previous work, developing failure prediction (fatigue) data for additional interconnect material choices, including aluminum and a variety of copper-Invar and copper-steel claddings. An improved global fatigue function is used to model the probability-of-failure statistics of each material as a function of level and number of cycles of applied strain. Life-cycle economic analyses are used to evaluate the relative merits of each material choce. The copper-Invar clad composites demonstrate superior performance over pure copper. Aluminum results are disappointing.

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.

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.

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

Numerical Analysis of Solids at Failure

DTIC Science & Technology

2011-08-20

failure analyses include the formulation of invariant finite elements for thin Kirchhoff rods, and preliminary initial studies of growth in...analysis of the failure of other structural/mechanical systems, including the finite element modeling of thin Kirchhoff rods and the constitutive...algorithm based on the connectivity graph of the underlying finite element mesh. In this setting, the discontinuities are defined by fronts propagating

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

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.

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

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

Electrical heating tests of uranium dioxide external fuel configuration at emitter temperature of 1900 K

NASA Technical Reports Server (NTRS)

Diianni, D. C.; Mayer, J. T.

1974-01-01

Testing of two fuel clad specimens for thermionic reactor application is described. The annular UO2 fuel was clad on both sides with tungsten; heat rejection was radially inward. The tests were intended to study inner clad stability, fuel redistribution, and fuel melting problems. The specimens were tested in a vacuum chamber using electron bombardment heating. Fuel structural changes were studied using periodic gammagraphs and posttest metallography. The first specimen test was terminated at 50 hours because of a braze failure. The second specimen was tested for 240 hours when an outer clad leak developed due to a tungsten-water reaction. The fuel developed numerous cracks on cooldown but the inner clad remained dimensionally stable. The fuel cover gas did not impede the rate of fuel redistribution. Posttest examination showed the fuel had not melted during operation.

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

The biomechanical consequences of rod reduction on pedicle screws: should it be avoided?

PubMed

Paik, Haines; Kang, Daniel G; Lehman, Ronald A; Gaume, Rachel E; Ambati, Divya V; Dmitriev, Anton E

2013-11-01

Rod contouring is frequently required to allow for appropriate alignment of pedicle screw-rod constructs. When residual mismatch is still present, a rod persuasion device is often used to achieve further rod reduction. Despite its popularity and widespread use, the biomechanical consequences of this technique have not been evaluated. To evaluate the biomechanical fixation strength of pedicle screws after attempted reduction of a rod-pedicle screw mismatch using a rod persuasion device. Fifteen 3-level, human cadaveric thoracic specimens were prepared and scanned for bone mineral density. Osteoporotic (n=6) and normal (n=9) specimens were instrumented with 5.0-mm-diameter pedicle screws; for each pair of comparison level tested, the bilateral screws were equal in length, and the screw length was determined by the thoracic level and size of the vertebra (35 to 45 mm). Titanium 5.5-mm rods were contoured and secured to the pedicle screws at the proximal and distal levels. For the middle segment, the rod on the right side was intentionally contoured to create a 5-mm residual gap between the inner bushing of the pedicle screw and the rod. A rod persuasion device was then used to engage the setscrew. The left side served as a control with perfect screw/rod alignment. After 30 minutes, constructs were disassembled and vertebrae individually potted. The implants were pulled in-line with the screw axis with peak pullout strength (POS) measured in Newton (N). For the proximal and distal segments, pedicle screws on the right side were taken out and reinserted through the same trajectory to simulate screw depth adjustment as an alternative to rod reduction. Pedicle screws reduced to the rod generated a 48% lower mean POS (495±379 N) relative to the controls (954±237 N) (p<.05) and significantly decreased work energy to failure (p<.05). Nearly half (n=7) of the pedicle screws had failed during the reduction attempt with visible pullout of the screw. After reduction, decreased POS was observed in both normal (p<.05) and osteoporotic (p<.05) bone. Back out and reinsertion of the screw resulted in no significant difference in mean POS, stiffness, and work energy to failure (p>.05). In circumstances where a rod is not fully seated within the pedicle screw, the use of a rod persuasion device decreases the overall POS and work energy to failure of the screw or results in outright failure. Further rod contouring or correction of pedicle screw depth of insertion may be warranted to allow for appropriate alignment of the longitudinal rods. Published by Elsevier Inc.

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 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

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.

Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

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

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

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

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

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

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.

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

Biomechanical stability according to different configurations of screws and rods.

PubMed

Ha, Kee-Yong; Hwang, Sung-Chul; Whang, Tae-Hyuk

2013-05-01

Comparison of biomechanical strength according to 2 different configurations of screws and rods. To compare the biomechanical strength of different configurations of screws and rods composed of the same material and of the same size. Many complications related to instrumentation have been reported. The incidence of metallic failure would differ according to the materials and configurations of the assembly of the screws and rods used. However, to our knowledge, the biomechanical effects of implant assembly rods and screws with different configurations and different contours have not been reported. Biomechanical testing was conducted to compare top tightening (TT) screw-rod configuration with side tightening (ST) screw-rod configuration. All tests were conducted using a hydraulic all-purpose testing machine. All data were acquired at a rate of 10 Hz. Both screw systems used spinal rods of 6 mm diameter and were made of TiAl4V ELI material. Among 5 types of tests, 3 were conducted on the basis of American Society for Testing and Materials (ASTM) F 1798 to 97 and F1717-10. The other 2 tests were conducted for comparing the characteristics between TT and ST pedicle screws according to modified methods from ASTM F 1717-10 and ASTM F 1798-97. All results including axial gripping capacity and yield forces were obtained using the same methods on the basis of the mentioned ASTM standards. In the axial gripping capacity test, the mean axial gripping capacity of the TT screw-rod configuration was 3332 ± 118 N and that of ST was 2222 ± 147 N in straight rods (P = 0.019). In 15-degree contoured rods, TT was 2988 ± 199 N and ST was 2116 ± 423 N (P = 0.014). In 30-degree contoured rods, TT was 2227 ± 408 N and ST was 1814 ± 285 N (P = 0.009). In the pulling-out test, the pulling-out force of ST was 8695 ± 1616 N and that of TT was 6106 ± 195 N (P = 0.014). In the rod-pushing test, the failure force of ST was 4131 ± 205 N and that of TT was 5639 ± 105 N. In the compressive fatigue test, the maximum load was 145 N in ST and 119 N in TT. In the cycle fatigue test, the fatigue strength of ST was higher than that of TT. In the rod-pushing test, the failure force of ST was 4131 ± 205 N and that of TT was 5639 ± 105 N (P=0.046). Two different configurations of rod-screw systems found statistically significant differences with axial gripping, pulling out, and fatigue failures. ST constructs improved fixation stability over TT constructs. It was concluded that ST configuration may reduce complications related to implantation.

Probabilistic analysis on the failure of reactivity control for the PWR

NASA Astrophysics Data System (ADS)

Sony Tjahyani, D. T.; Deswandri; Sunaryo, G. R.

2018-02-01

The fundamental safety function of the power reactor is to control reactivity, to remove heat from the reactor, and to confine radioactive material. The safety analysis is used to ensure that each parameter is fulfilled during the design and is done by deterministic and probabilistic method. The analysis of reactivity control is important to be done because it will affect the other of fundamental safety functions. The purpose of this research is to determine the failure probability of the reactivity control and its failure contribution on a PWR design. The analysis is carried out by determining intermediate events, which cause the failure of reactivity control. Furthermore, the basic event is determined by deductive method using the fault tree analysis. The AP1000 is used as the object of research. The probability data of component failure or human error, which is used in the analysis, is collected from IAEA, Westinghouse, NRC and other published documents. The results show that there are six intermediate events, which can cause the failure of the reactivity control. These intermediate events are uncontrolled rod bank withdrawal at low power or full power, malfunction of boron dilution, misalignment of control rod withdrawal, malfunction of improper position of fuel assembly and ejection of control rod. The failure probability of reactivity control is 1.49E-03 per year. The causes of failures which are affected by human factor are boron dilution, misalignment of control rod withdrawal and malfunction of improper position for fuel assembly. Based on the assessment, it is concluded that the failure probability of reactivity control on the PWR is still within the IAEA criteria.

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

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

Failure kinetics in borosilicate glass during rod impact

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

Orphal, Dennis L.; Anderson, Charles E. Jr.; Behner, Thilo

2007-12-12

Failure front (FF) and penetration velocity have been measured for long gold rods impacting and penetrating borosilicate (BS) glass. Data are obtained by visualizing simultaneously FF propagation with a high speed camera and rod penetration with flash X-rays. Results for BS glass are qualitatively similar to those of DEDF (PbO) glass. FF velocity rapidly decreases from an initial value to a lower, approximately constant value. FF velocity increases with impact velocity, v{sub p}. The FF velocity remains significantly lower than the shear velocity, even at the highest impact velocity tested, about 2.5 km/s. The ratio of the FF velocity tomore » the rod penetration velocity, v{sub F}/u, decreases with increasing v{sub p} and appears to be approaching v{sub F}/u = 1 asymptotically, as observed previously for DEDF glass. The separation of the FF and the tip of the rod decreases with increasing impact velocity. Importantly, since v{sub F}/u{>=}1, the gold rod is always penetrating glass behind the FF.« less

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

Failure Wave in DEDF and Soda-Lime Glass During Rod Impact

NASA Astrophysics Data System (ADS)

Orphal, Dennis; Behner, Thilo; Anderson, Charles; Templeton, Douglas

2005-07-01

Investigations of glass by planar, and classical and symmetric Taylor impact experiments reveal that failure wave velocity U/F depends on impact velocity, geometry, and the type of glass. U/F typically increases with impact velocity to between ˜ 1.4 C/S and C/L (shear and longitudinal wave velocities, respectively). This paper reports the results of direct high-speed photographic measurements of the failure wave for gold rod impact from 1.2 and 2.0 km/s on DEDF glass (C/S = 2.0, C/L =3.5 km/s). The average rod penetration velocity, u, was measured using flash X-rays. Gold rods eliminated penetrator strength effects. U/F for gold rod impact on DEDF is ˜ 1.0-1.2 km/s, which is considerably less than C/S. The increase of u with impact velocity is greater than that of U/F. These results are confirmed by soda-lime glass impact on a gold rod at an impact velocity of 1300 m/s. Similar results are found in``edge-on-impact'' tests; U/F values of 1.4 km/s and 2.4-2.6 km/s in soda-lime glass are reported for W-alloy rod impact, considerably less than C/S (3.2 km/s) [1,2]. [1] Bless, et. al.(1990) AIP Proc. Shock Comp. Cond. Matter---1989, pp. 939-942 (1990) [2] E. L. Zilberbrand, et. al. (1999) Int. J. Impact Engng., 23, 995-1001 (1999).

76 FR 64798 - Airworthiness Directives; Airbus Model A300 B4-103, B4-203, and B4-2C Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-19

... condition as: One operator reported a failure of the MLG [main landing gear] retraction actuator sliding rod.... The MCAI states: One operator reported a failure of the MLG [main landing gear] retraction actuator..., December 10, 2007)] to require repetitive inspections of the retraction actuator sliding rod as installed...

Microstructure and mechanical properties of hot wire laser clad layers for repairing precipitation hardening martensitic stainless steel

NASA Astrophysics Data System (ADS)

Wen, Peng; Cai, Zhipeng; Feng, Zhenhua; Wang, Gang

2015-12-01

Precipitation hardening martensitic stainless steel (PH-MSS) is widely used as load-bearing parts because of its excellent overall properties. It is economical and flexible to repair the failure parts instead of changing new ones. However, it is difficult to keep properties of repaired part as good as those of the substrate. With preheating wire by resistance heat, hot wire laser cladding owns both merits of low heat input and high deposition efficiency, thus is regarded as an advantaged repairing technology for damaged parts of high value. Multi-pass layers were cladded on the surface of FV520B by hot wire laser cladding. The microstructure and mechanical properties were compared and analyzed for the substrate and the clad layer. For the as-cladded layer, microstructure was found non-uniform and divided into quenched and tempered regions. Tensile strength was almost equivalent to that of the substrate, while ductility and impact toughness deteriorated much. With using laser scanning layer by layer during laser cladding, microstructure of the clad layers was tempered to fine martensite uniformly. The ductility and toughness of the clad layer were improved to be equivalent to those of the substrate, while the tensile strength was a little lower than that of the substrate. By adding TiC nanoparticles as well as laser scanning, the precipitation strengthening effect was improved and the structure was refined in the clad layer. The strength, ductility and toughness were all improved further. Finally, high quality clad layers were obtained with equivalent or even superior mechanical properties to the substrate, offering a valuable technique to repair PH-MSS.

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

Micromechanics of failure waves in glass. 2: Modeling

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

Espinosa, H.D.; Xu, Y.; Brar, N.S.

1997-08-01

In an attempt to elucidate the failure mechanism responsible for the so-called failure waves in glass, numerical simulations of plate and rod impact experiments, with a multiple-plane model, have been performed. These simulations show that the failure wave phenomenon can be modeled by the nucleation and growth of penny-shaped shear defects from the specimen surface to its interior. Lateral stress increase, reduction of spall strength,and progressive attenuation of axial stress behind the failure front are properly predicted by the multiple-plane model. Numerical simulations of high-strain-rate pressure-shear experiments indicate that the model predicts reasonably well the shear resistance of the materialmore » at strain rates as high as 1 {times} 10{sup 6}/s. The agreement is believed to be the result of the model capability in simulating damage-induced anisotropy. By examining the kinetics of the failure process in plate experiments, the authors show that the progressive glass spallation in the vicinity of the failure front and the rate of increase in lateral stress are more consistent with a representation of inelasticity based on shear-activated flow surfaces, inhomogeneous flow, and microcracking, rather than pure microcracking. In the former mechanism, microcracks are likely formed at a later time at the intersection of flow surfaces, in the case of rod-on-rod impact, stress and radial velocity histories predicted by the microcracking model are in agreement with the experimental measurements. Stress attenuation, pulse duration, and release structure are properly simulated. It is shown that failure wave speeds in excess to 3,600 m/s are required for adequate prediction in rod radial expansion.« less

Control rod drive

DOEpatents

Hawke, Basil C.

1986-01-01

A control rod drive uses gravitational forces to insert one or more control rods upwardly into a reactor core from beneath the reactor core under emergency conditions. The preferred control rod drive includes a vertically movable weight and a mechanism operatively associating the weight with the control rod so that downward movement of the weight is translated into upward movement of the control rod. The preferred control rod drive further includes an electric motor for driving the control rods under normal conditions, an electrically actuated clutch which automatically disengages the motor during a power failure and a decelerator for bringing the control rod to a controlled stop when it is inserted under emergency conditions into a reactor core.

77 FR 46940 - Airworthiness Directives; Glasflugel Gliders

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-07

... condition as corrosion damage to the elevator control rod that could lead to failure of the elevator control... into the elevator control rod through a control bore hole and resulted in corrosion damage. The investigation concluded as well that the corrosion cannot be detected from outside the elevator control rod...

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.

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.

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.

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.

Failure Wave in DEDF and Soda-Lime Glass during Rod Impact

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

Orphal, D. L.; Behner, Th.; Hohler, V.

2006-07-28

Investigations of glass by planar, and classical and symmetric Taylor impact experiments reveal that failure wave velocity vF depends on impact velocity, geometry, and type of glass. vF typically increases with impact velocity vP to between cS and cL or to {radical}2cS (shear and longitudinal wave velocity). This paper reports initial results of an investigation of failure waves associated with gold rod impact on high-density (DEDF) glass and soda-lime glass. Data are obtained by visualizing simultaneously the failure propagation in the glass with a high-speed camera and the rod penetration velocity u with flash radiography. Results for DEDF glass aremore » reported for vP between 1.2 and 2.0 km/s, those for soda-lime glass with vP {approx_equal}1.3 km/s. It is shown that vF > u, and that in the case of DEDF glass vF/u decreases from ; 1.38 to 1.13 with increasing vp. In addition, several Taylor tests were performed. For both DEDF and soda-lime glass the vF-values, found here as well as vF- data reported in the literature, reveal that--for equal pressures--the failure wave velocities determined from Taylor tests or planar-impact tests are distinctly greater than those observed during steady-state rod penetration.« less

Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

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

Clarno, Kevin T.

2016-01-30

The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, andmore » at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.« less

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

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.

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.

Detonation Failure Thickness Measurement in AN Annular Geometry

NASA Astrophysics Data System (ADS)

Mack, D. B.; Petel, O. E.; Higgins, A. J.

2007-12-01

The failure thickness of neat nitromethane in aluminum confinement was measured using a novel experimental technique. The thickness was approximated in an annular geometry by the gap between a concentric aluminum tube and rod. This technique was motivated by the desire to have a periodic boundary condition in the direction orthogonal to the annulus thickness, rather than a free surface occurring in typical rectangular geometry experiments. This results in a two-dimensional charge analogous to previous failure thickness setups but with infinite effective width (i.e. infinite aspect ratio). Detonation propagation or failure was determined by the observation of failure patterns engraved on the aluminum rod by the passing detonation. Analysis of these engraved patterns provides a statistical measurement of the spatial density of failure waves. Failure was observed as far as 180 thicknesses downstream. The failure thickness was measured to be 1.45 mm±0.15 mm.

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

Fastening of a High-Strength Composite rod with a Splitted and Wedged end in a Potted Anchor 1. Experimental Investigation

NASA Astrophysics Data System (ADS)

Arnautov, A. K.; Terrasi, G. P.; Kulakov, V. L.; Portnov, G. G.

2014-01-01

The effectiveness of fastening of high-strength unidirectional CFRP/epoxy rods in potted anchors was investigated experimentally. The rods had splitted ends, in which duralumin wedges were glued. The experiments, performed for three types of contact between the composite rods and the potted material, showed that the most effective were full adhesion and adhesion-friction contacts, when the maximum load-carrying capacity of CFRP rods under tension could be reached. The full friction contact was ineffective because of the shear failure of CFRP rods inside the anchorage zone.

A Jacobian-free Newton Krylov method for mortar-discretized thermomechanical contact problems

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

Hansen, Glen, E-mail: Glen.Hansen@inl.gov

2011-07-20

Multibody contact problems are common within the field of multiphysics simulation. Applications involving thermomechanical contact scenarios are also quite prevalent. Such problems can be challenging to solve due to the likelihood of thermal expansion affecting contact geometry which, in turn, can change the thermal behavior of the components being analyzed. This paper explores a simple model of a light water reactor nuclear fuel rod, which consists of cylindrical pellets of uranium dioxide (UO{sub 2}) fuel sealed within a Zircalloy cladding tube. The tube is initially filled with helium gas, which fills the gap between the pellets and cladding tube. Themore » accurate modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel performance, including cladding stress and behavior under irradiated conditions, which are factors that affect the lifetime of the fuel. The thermomechanical contact approach developed here is based on the mortar finite element method, where Lagrange multipliers are used to enforce weak continuity constraints at participating interfaces. In this formulation, the heat equation couples to linear mechanics through a thermal expansion term. Lagrange multipliers are used to formulate the continuity constraints for both heat flux and interface traction at contact interfaces. The resulting system of nonlinear algebraic equations are cast in residual form for solution of the transient problem. A Jacobian-free Newton Krylov method is used to provide for fully-coupled solution of the coupled thermal contact and heat equations.« less

A Jacobian-Free Newton Krylov Method for Mortar-Discretized Thermomechanical Contact Problems

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

Glen Hansen

2011-07-01

Multibody contact problems are common within the field of multiphysics simulation. Applications involving thermomechanical contact scenarios are also quite prevalent. Such problems can be challenging to solve due to the likelihood of thermal expansion affecting contact geometry which, in turn, can change the thermal behavior of the components being analyzed. This paper explores a simple model of a light water reactor nuclear reactor fuel rod, which consists of cylindrical pellets of uranium dioxide (UO2) fuel sealed within a Zircalloy cladding tube. The tube is initially filled with helium gas, which fills the gap between the pellets and cladding tube. Themore » accurate modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel performance, including cladding stress and behavior under irradiated conditions, which are factors that affect the lifetime of the fuel. The thermomechanical contact approach developed here is based on the mortar finite element method, where Lagrange multipliers are used to enforce weak continuity constraints at participating interfaces. In this formulation, the heat equation couples to linear mechanics through a thermal expansion term. Lagrange multipliers are used to formulate the continuity constraints for both heat flux and interface traction at contact interfaces. The resulting system of nonlinear algebraic equations are cast in residual form for solution of the transient problem. A Jacobian-free Newton Krylov method is used to provide for fully-coupled solution of the coupled thermal contact and heat equations.« 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

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.

Superflywheel energy storage system. [for windpowered machines

NASA Technical Reports Server (NTRS)

Rabenhorst, D. W.

1973-01-01

A windpowered system using the superflywheel configuration for energy storage is considered. Basic elements of superflywheels are thin rods assembled in pregrooved hub lamina so that they fan out in radial orientation. Adjacent layers of hub lamina are assembled 90 degree in rotation to each other so as to form a circular brush configuration. Thus stress concentrations and rod failure are minimized and realistic failure containment for a high performance flywheel is obtained.

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

CONTROL ROD DRIVE

DOEpatents

Chapellier, R.A.

1960-05-24

BS>A drive mechanism was invented for the control rod of a nuclear reactor. Power is provided by an electric motor and an outside source of fluid pressure is utilized in conjunction with the fluid pressure within the reactor to balance the loadings on the motor. The force exerted on the drive mechanism in the direction of scramming the rod is derived from the reactor fluid pressure so that failure of the outside pressure source will cause prompt scramming of the rod.

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Single- versus dual-rod anterior instrumentation of thoracolumbar curves in adolescent idiopathic scoliosis.

PubMed

Nambiar, Mithun; Yang, Yi; Liew, Susan; Turner, Peter L; Torode, Ian P

2016-10-01

Single or dual-rod instrumentation can be used for the anterior fixation of the spine in adolescent idiopathic scoliosis (AIS). We aim to compare the complications, radiographic and functional outcomes of patients with AIS who have undergone single and dual-rod instrumentation. This is a multi-centre study involving the Royal Children's, Royal Melbourne and Epworth hospitals. Three primary surgeons were involved to ensure homogeneity of surgical technique and implants. Patients with AIS and thoracolumbar curves (Lenke 5 and 6) undergoing anterior instrumentation from 1st January 2000 to 30th June 2013 were included. Radiographic data were collected from X-rays. The functional outcome was measured through the Scoliosis Research Society questionnaire (SRS-30). The study included 58 patients (38 single-rod and 20 dual-rod patients). Thirty-nine patients were classified with Lenke 5 curves, while 19 patients had Lenke 6 curves. Structural interbody supports were used in 95 % of cases. In the preoperative to postoperative period, patients with single rods had an improvement of 75 and 51 % for primary and secondary curves, respectively, while patients with dual rods had an improvement of 70 and 38 % for primary and secondary curves, respectively. There were no cases of pseudoarthrosis or metalware failure in either group. Two patients (one single-rod and one dual-rod patient) required further unplanned posterior fusion. 91 % of patients were satisfied with the results of their back management. Pseudoarthrosis and metalware failure are rare complications of anterior instrumentation. Our study found no significant difference in functional or radiographic outcome between single and dual-rod instrumentation. Level III.

Performance evaluation of a semi-active cladding connection for multi-hazard mitigation

NASA Astrophysics Data System (ADS)

Gong, Yongqiang; Cao, Liang; Micheli, Laura; Laflamme, Simon; Quiel, Spencer; Ricles, James

2018-03-01

A novel semi-active damping device termed Variable Friction Cladding Connection (VFCC) has been previously proposed to leverage cladding systems for the mitigation of natural and man-made hazards. The VFCC is a semi-active friction damper that connects cladding elements to the structural system. The friction force is generated by sliding plates and varied using an actuator through a system of adjustable toggles. The dynamics of the device has been previously characterized in a laboratory environment. In this paper, the performance of the VFCC at mitigating non-simultaneous multi-hazard excitations that includes wind and seismic loads is investigated on a simulated benchmark building. Simulations consider the robustness with respect to some uncertainties, including the wear of the friction surfaces and sensor failure. The performance of the VFCC is compared against other connection strategies including traditional stiffness, passive viscous, and passive friction elements. Results show that the VFCC is robust and capable of outperforming passive systems for the mitigation of multiple hazards.

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

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.

Mechanical beam isolator for high-power laser systems

DOEpatents

Post, Richard F.; Vann, Charles S.

1998-01-01

A mechanical beam isolator uses rod-shaped elements having a Gaussian configuration to interrupt the path of a beam of photons or particles when the time-scale of the needed interruption is of the order of a microsecond or less. One or more of these rods is mounted transversely to, and penetrates through, a rotating shaft supported by bearings. Owing to the Gaussian geometry of the rods, they are able to withstand much higher rotation speeds, without tensile failure, than rods having any other geometrical shape.

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

Posttest examination results of recent treat tests on metal fuel

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

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

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

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.

1973-01-01

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

High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

NASA Astrophysics Data System (ADS)

Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

2014-08-01

In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover, for as high as 1600 °C, the non-reactive LM was experimentally confirmed not to show any chemical reaction with air or moisture in the air. This experimental work confirmed the excellent compatibility behaviors between the LM as a PWR fuel gap filler and stainless steel and high-density concrete in the high-temperature regime.

Sensitivity to VSR failure: K pipe break accident

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

Meichle, R.H.

1969-09-12

Reactor effects of failure of a safety rod to scram can be considered in two major respects: The reduction in total safety system strength which will affect the amount of ``prompt drop`` and subsequent flux decay rate of the average neutron flux-level; and the change in local flux distribution due to the absence of the particular rod which fails to enter the reactor. The purpose of this memorandum is to describe the physical effects involved and to indicate the approximate magnitude of both reactor-wide and localized changes in event of failure of a VSR simultaneous with a K Reactor risermore » accident.« less

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

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

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

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

Fuel inspection and reconstitution experience at Surry Power Station

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

Brookmire, T.A.

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

Nuclear fuel performance: Trends, remedies and challenges

NASA Astrophysics Data System (ADS)

Rusch, C. A.

2008-12-01

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

Electrochemical Migration Behavior of Copper-Clad Laminate and Electroless Nickel/Immersion Gold Printed Circuit Boards under Thin Electrolyte Layers

PubMed Central

Yi, Pan; Xiao, Kui; Ding, Kangkang; Dong, Chaofang; Li, Xiaogang

2017-01-01

The electrochemical migration (ECM) behavior of copper-clad laminate (PCB-Cu) and electroless nickel/immersion gold printed circuit boards (PCB-ENIG) under thin electrolyte layers of different thicknesses containing 0.1 M Na2SO4 was studied. Results showed that, under the bias voltage of 12 V, the reverse migration of ions occurred. For PCB-Cu, both copper dendrites and sulfate precipitates were found on the surface of FR-4 (board material) between two plates. Moreover, the Cu dendrite was produced between the two plates and migrated toward cathode. Compared to PCB-Cu, PCB-ENIG exhibited a higher tendency of ECM failure and suffered from seriously short circuit failure under high relative humidity (RH) environment. SKP results demonstrated that surface potentials of the anode plates were greater than those of the cathode plates, and those potentials of the two plates exhibited a descending trend as the RH increased. At the end of the paper, an electrochemical migration corrosion failure model of PCB was proposed. PMID:28772497

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.

Mechanical beam isolator for high-power laser systems

DOEpatents

Post, R.F.; Vann, C.S.

1998-07-07

A mechanical beam isolator uses rod-shaped elements having a Gaussian configuration to interrupt the path of a beam of photons or particles when the time-scale of the needed interruption is of the order of a microsecond or less. One or more of these rods is mounted transversely to, and penetrates through, a rotating shaft supported by bearings. Owing to the Gaussian geometry of the rods, they are able to withstand much higher rotation speeds, without tensile failure, than rods having any other geometrical shape. 3 figs.

Severe accident modeling of a PWR core with different cladding materials

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

Johnson, S. C.; Henry, R. E.; Paik, C. Y.

2012-07-01

The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCSmore » rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)« less

Oral manifestations in a renal osteodystrophy patient - a case report with review of literature.

PubMed

J, Parthiban; Nisha V, Aarthi; Gs, Asokan; Ca, Prakash; Mm, Varadharaja

2014-08-01

Renal Osteodystrophy (ROD) is a common complication of chronic renal disease (CRD) and is the part of a broad spectrum of disorders of mineral metabolism that occurs in the clinical setting. It occurs early in the course of chronic renal failure and progresses as the kidney function deteriorates. It is an osseous alteration believed to arise from increased parathyroid function associated with inappropriate calcium, phosphorus and vitamin D metabolism. Involvement of the jaws is common and radiographic alterations are often one of the earliest signs of chronic renal failure. Herein, reporting a case of Chronic Renal Failure (Bilateral Grade I Neuropathy) with ROD presenting oral manifestations in an 11-year -old male child.

Competing mechanisms in the wear resistance behavior of biomineralized rod-like microstructures

NASA Astrophysics Data System (ADS)

Escobar de Obaldia, Enrique; Herrera, Steven; Grunenfelder, Lessa Kay; Kisailus, David; Zavattieri, Pablo

2016-11-01

The remarkable mechanical properties observed in biological composite materials relative to those of their individual constituents distinguish them from common engineering materials. Some naturally occurring high-performance ceramics, like the external veneer of the Chiton (Cryptochiton stelleri) tooth, have been shown to have superior hardness and impressive abrasion resistance properties. The mechanical performance of the chiton tooth has been attributed to a hierarchical arrangement of nanostructured magnetite rods surrounded with organic material. While nanoindentation tests provide useful information about the overall performance of this biological composite, understanding the key microstructural features and energy dissipation mechanisms at small scales remains a challenging task. We present a combined experimental/numerical approach to elucidate the role of material deformation in the rods, debonding at the rod interfaces and the influence of energy dissipation mechanisms on the ability of the microstructure to distribute damage under extreme loading conditions. We employ a 3D finite element-based micromechanical model to simulate the nanoindentation tests performed in geological magnetite and cross-sections of the chiton tooth. This proposed model is capable of capturing the inelastic deformation of the rods and the failure of their interfaces, while damage, fracture and fragmentation of the mineralized rods is assessed using a probabilistic function. Our results show that these natural materials achieve their abrasion resistant properties by controlling the interface strength between rods, alleviating the tensile stress on the rods near the indentation tip and therefore decreasing the probability of catastrophic failure without significantly sacrificing resistance to penetration. The understanding of these competing energy dissipating mechanisms provides a path to the prediction of new combination of materials. In turns, these results suggest certain guidelines for abrasion resistance rod-like microstructures in composites with high volume fraction of brittle minerals or ceramics with tailored performance for specific applications.

Multifrequency Eddy Current Inspection of Corrosion in Clad Aluminum Riveted Lap Joints and Its Effect on Fatigue Life

NASA Astrophysics Data System (ADS)

Okafor, A. C.; Natarajan, S.

2007-03-01

Aging aircraft are prone to corrosion damage and fatigue cracks in riveted lap joints of fuselage skin panels. This can cause catastrophic failure if not detected and repaired. Hence detection of corrosion damage and monitoring its effect on structural integrity are essential. This paper presents multifrequency eddy current (EC) inspection of corrosion damage and machined material loss defect in clad A1 2024-T3 riveted lap joints and its effect on fatigue life. Results of eddy current inspection, corrosion product removal and fatigue testing are presented.

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.

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

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

DOE PAGES

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

2016-02-27

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

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

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

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

2016-05-01

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

Effect of screw position on single cycle to failure in bending and torsion of a locking plate-rod construct in a synthetic feline femoral gap model.

PubMed

Niederhäuser, Simone K; Tepic, Slobodan; Weber, Urs T

2015-05-01

To evaluate the effect of screw position on strength and stiffness of a combination locking plate-rod construct in a synthetic feline femoral gap model. 30 synthetic long-bone models derived from beechwood and balsa wood. 3 constructs (2 locking plate-rod constructs and 1 locking plate construct; 10 specimens/construct) were tested in a diaphyseal bridge plating configuration by use of 4-point bending and torsion. Variables included screw position (near the fracture gap and far from the fracture gap) and application of an intramedullary pin. Constructs were tested to failure in each loading mode to determine strength and stiffness. Failure was defined as plastic deformation of the plate or breakage of the bone model or plate. Strength, yield angle, and stiffness were compared by use of a Wilcoxon test. Placement of screws near the fracture gap did not increase bending or torsional stiffness in the locking plate-rod constructs, assuming the plate was placed on the tension side of the bone. Addition of an intramedullary pin resulted in a significant increase in bending strength of the construct. Screw positioning did not have a significant effect on any torsion variables. Results of this study suggested that, in the investigated plate-rod construct, screw insertion adjacent to the fracture lacked mechanical advantages over screw insertion at the plate ends. For surgeons attempting to minimize soft tissue dissection, the decision to make additional incisions for screw placement should be considered with even more caution.

Oral Manifestations in a Renal Osteodystrophy Patient - A Case Report with Review of Literature

PubMed Central

Nisha V, Aarthi; GS, Asokan; CA, Prakash; MM, Varadharaja

2014-01-01

Renal Osteodystrophy (ROD) is a common complication of chronic renal disease (CRD) and is the part of a broad spectrum of disorders of mineral metabolism that occurs in the clinical setting. It occurs early in the course of chronic renal failure and progresses as the kidney function deteriorates. It is an osseous alteration believed to arise from increased parathyroid function associated with inappropriate calcium, phosphorus and vitamin D metabolism. Involvement of the jaws is common and radiographic alterations are often one of the earliest signs of chronic renal failure. Herein, reporting a case of Chronic Renal Failure (Bilateral Grade I Neuropathy) with ROD presenting oral manifestations in an 11-year -old male child. PMID:25302278

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

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.

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

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

Field, J.H.

1962-11-15

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

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

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

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.

10 CFR 55.59 - Requalification.

Code of Federal Regulations, 2012 CFR

2012-01-01

... system. (U) Fuel cladding failure or high activity in reactor coolant or offgas. (V) Turbine or generator... heatup rate is established. (B) Plant shutdown. (C) Manual control of steam generators or feedwater or... generator leaks (2)Inside and outside primary containment (3)Large and small, including lead-rate...

10 CFR 55.59 - Requalification.

Code of Federal Regulations, 2011 CFR

2011-01-01

... system. (U) Fuel cladding failure or high activity in reactor coolant or offgas. (V) Turbine or generator... heatup rate is established. (B) Plant shutdown. (C) Manual control of steam generators or feedwater or... generator leaks (2)Inside and outside primary containment (3)Large and small, including lead-rate...

10 CFR 55.59 - Requalification.

Code of Federal Regulations, 2013 CFR

2013-01-01

... system. (U) Fuel cladding failure or high activity in reactor coolant or offgas. (V) Turbine or generator... heatup rate is established. (B) Plant shutdown. (C) Manual control of steam generators or feedwater or... generator leaks (2)Inside and outside primary containment (3)Large and small, including lead-rate...

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

Status of Wrought FeCrAl-UO 2 Capsules Irradiated in the Advanced Test Reactor

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

Field, Kevin G.; Harp, J.; Core, G.

2017-07-01

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction,more » and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.« less

Vibrations of beams and rods carrying a moving mass

NASA Astrophysics Data System (ADS)

Zhao, X. W.; van der Heijden, G. H. M.; Hu, Z. D.

2016-05-01

We study the vibration of slender one-dimensional elastic structures (beams, cables, wires, rods) under the effect of a moving mass or load. We first consider the classical small- deflection (Euler-Bernoulli) beam case, where we look at tip vibrations of a cantilever as a model for a barreled launch system. Then we develop a theory for large deformations based on Cosserat rod theory. We illustrate the effect of moving loads on large-deformation structures with a few cable and arch problems. Large deformations are found to have a resonance detuning effect on the cable. For the arch we find different failure modes depending on its depth: a shallow arch fails by in-plane collapse, while a deep arch fails by sideways flopping. In both cases the speed of the traversing load is found to have a stabilising effect on the structure, with failure suppressed entirely at sufficiently high speed.

Bone Plating in Patients with Type III Osteogenesis Imperfecta: Results and Complications

PubMed Central

Enright, William J; Noonan, Kenneth J

2006-01-01

The results of bone plating in four children (6 femurs, 2 tibias) with osteogenesis imperfecta type III were analyzed. Average age at time of operation was 44 months. In three of the femurs, multiple platings were performed for a total of 13 bone platings in the eight bones studied. Average time to revision following plating was 27 months. Indications for revision included fracture (6), deformity (3), hardware failure (3), and nonunion (1). Other complications included one case of compartment syndrome. All eight bones were ultimately revised to elongating intramedullary Bailey-Dubow rods. Bone plating in skeletally immature patients with osteogenesis imperfecta does not provide better outcome than elongating rods. Complications from bone plating leading to revision, such as refracture or hardware failure, are higher than in those children managed with elongating rods, as previously reported in the literature. PMID:16789446

Simulation Studies of Mechanical Properties of Novel Silica Nano-structures

NASA Astrophysics Data System (ADS)

Muralidharan, Krishna; Torras Costa, Joan; Trickey, Samuel B.

2006-03-01

Advances in nanotechnology and the importance of silica as a technological material continue to stimulate computational study of the properties of possible novel silica nanostructures. Thus we have done classical molecular dynamics (MD) and multi-scale quantum mechanical (QM/MD) simulation studies of the mechanical properties of single-wall and multi-wall silica nano-rods of varying dimensions. Such nano-rods have been predicted by Mallik et al. to be unusually strong in tensile failure. Here we compare failure mechanisms of such nano-rods under tension, compression, and bending. The concurrent multi-scale QM/MD studies use the general PUPIL system (Torras et al.). In this case, PUPIL provides automated interoperation of the MNDO Transfer Hamiltonian QM code (Taylor et al.) and a locally written MD code. Embedding of the QM-forces domain is via the scheme of Mallik et al. Work supported by NSF ITR award DMR-0325553.

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2011 CFR

2011-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2010 CFR

2010-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2013 CFR

2013-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2012 CFR

2012-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2014 CFR

2014-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

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.

Microstructural and Material Quality Effects on Rolling Contact Fatigue of Highly Elastic Intermetallic Ball Bearings

NASA Technical Reports Server (NTRS)

DellaCorte, Christopher; Howard, S. Adam; Thomas, Fransua; Stanford, Malcolm K.

2016-01-01

Rolling element bearings made from highly-elastic intermetallic materials (HIM)s, such as 60NiTi, are under development for applications that require superior corrosion and shock resistance. Compared to steel, intermetallics have been shown to have much lower rolling contact fatigue (RCF) stress capability in simplified 3-ball on rod (ASTM STP 771) fatigue tests. In the 3-ball tests, poor material quality and microstructural flaws negatively affect fatigue life but such relationships have not been established for full-scale 60NiTi bearings. In this paper, 3-ball-on-rod fatigue behavior of two quality grades of 60NiTi are compared to the fatigue life of full-scale 50mm bore ball bearings made from the same materials. 60NiTi RCF rods with material or microstructural flaws suffered from infant mortality failures at all tested stress levels while high quality 60NiTi rods exhibited no failures at lower stress levels. Similarly, tests of full-scale bearings made from flawed materials exhibited early surface fatigue and through crack type failures while bearings made from high quality material did not fail even in long-term tests. Though the full-scale bearing test data is yet preliminary, the results suggest that the simplified RCF test is a good qualitative predictor of bearing performance. These results provide guidance for materials development and to establish minimum quality levels required for successful bearing operation and life.

Microstructural and Material Quality Effects on Rolling Contact Fatigue of Highly Elastic Intermetallic NiTi Ball Bearings

NASA Technical Reports Server (NTRS)

Dellacorte, Christopher; Howard, S. Adam; Thomas, Fransua; Stanford, Malcolm K.

2017-01-01

Rolling element bearings made from highly-elastic intermetallic materials (HIM)s, such as 60NiTi, are under development for applications that require superior corrosion and shock resistance. Compared to steel, intermetallics have been shown to have much lower rolling contact fatigue (RCF) stress capability in simplified 3-ball on rod (ASTM STP 771) fatigue tests. In the 3-ball tests, poor material quality and microstructural flaws negatively affect fatigue life but such relationships have not been established for full-scale 60NiTi bearings. In this paper, 3-ball-on-rod fatigue behavior of two quality grades of 60NiTi are compared to the fatigue life of full-scale 50mm bore ball bearings made from the same materials. 60NiTi RCF rods with material or microstructural flaws suffered from infant mortality failures at all tested stress levels while high quality 60NiTi rods exhibited no failures at lower stress levels. Similarly, tests of full-scale bearings made from flawed materials exhibited early surface fatigue and through crack type failures while bearings made from high quality material did not fail even in long-term tests. Though the full-scale bearing test data is yet preliminary, the results suggest that the simplified RCF test is a good qualitative predictor of bearing performance. These results provide guidance for materials development and to establish minimum quality levels required for successful bearing operation and life.

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

DEVICE FOR CONTROLLING INSERTION OF ROD

DOEpatents

Beaty, B.J.

1958-10-14

A device for rapidly inserting a safety rod into a nuclear reactor upon a given signal or in the event of a power failure in order to prevent the possibility of extensive damage caused by a power excursion is described. A piston is slidably mounted within a vertical cylinder with provision for an electromagnetic latch at the top of the cylinder. This assembly, with a safety rod attached to the piston, is mounted over an access port to the core region of the reactor. The piston is normally latched at the top of the cylinder with the safety rod clear of the core area, however, when the latch is released, the piston and rod drop by their own weight to insert the rod. Vents along the side of the cylinder permit the escape of the air entrapped under the piston over the greater part of the distance, however, at the end of the fall the entrapped air is compressed thereby bringing the safety rod gently to rest, thus providing for a rapid automatic insertion of the rod with a minimum of structural shock.

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

NASA Astrophysics Data System (ADS)

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

2016-05-01

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

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

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

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

1961-08-01

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

75 FR 68548 - Airworthiness Directives; Airbus Model A318, A319, A320, and A321 Series Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-08

...: One case of elevator servo-control disconnection has been experienced on an aeroplane of the A320 family. Investigation has revealed that the failure occurred at the servo-control rod eye-end. Further to... servo-control rod eye-ends. In several cases, both actuators of the same elevator surface were affected...

Design test request No. 1263 K Reactor graphite key and VSR channel sleeve test

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

Kempf, F.J.

1964-12-10

The objectives of this test were: (1) Determine the coefficient of friction between two adjacent layers of K Reactor graphite at room temperature. (2) Determine the average load required to cause failure of an unirradiated K Reactor side reflector bar, when subjected to tensile loading applied through the reflector keys. (3) Determine the average load at failure and the average deflection at failure of a single VSR channel key when loaded in keyways with clearances equal to those used in original stack construction. (4) Determine the average load and deflection required to break the four K Reactor VSR keys whenmore » loaded simultaneously in both `3-layer` and `7-layer` mockups. Also determine the mode of key failure; i.e., shear, flexure or combined compression and bending. Following these key rupture tests, determine the strength and deflection characteristics of the proposed K Reactor VSR channel sleeve when loaded in a manner identical to that used to fracture the keys. (5) Determine the average load and deflection at failure of both the proposed K Reactor VSR channel sleeves and the proposed C Reactor sleeves when subjected to crushing loads. (6) Determine the extent of damage to the proposed K Reactor VSR channel sleeve when subjected to the following vertical rod loading conditions. (a) Full rod drop in a channel mockup which has been misaligned 2 1/2 inches. (b) Full rod drop in a channel which has been misaligned an amount equal to the maximum flexibility of a `universal` VSR.« less

QUICK RELEASABLE DRIVE

DOEpatents

Dickson, J.J.

1958-07-01

A quick releasable mechanical drive system suitable for use in a nuclear reactor is described. A small reversible motor positions a control rod by means of a worm and gear speed reducer, a magnetic torque clutch, and a bell crank. As the control rod is raised to the operating position, a heavy coil spring is compressed. In the event of an emergency indicated by either a''scram'' signal or a power failure, the current to the magnetic clutch is cut off, thereby freeing the coil spring and the bell crank positioner from the motor and speed reduction gearing. The coil spring will immediately act upon the bell crank to cause the insertion of the control rod. This arrangement will allow the slow, accurate positioning of the control rod during reactor operation, while providing an independent force to rapidly insert the rod in the event of an emergency.

In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

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

Karb, E.H.

1980-01-01

In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, whichmore » has been observed in all tests with previously irradiated rods, needs further investigation.« less

The Acoustic Emission signal acquired by the microphones placed in the CABRI test device along the fourteen last R.I.A. experiments: an example of reproducible research in nuclear science

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

Laurent Pantera, Oumar Traore

The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission. It is located at the Cadarache Research Centre in southern France and it is designed to act as a support to the French nuclear infrastructure. The purpose of the new testing programme termed, 'CABRI International Programme' (CIP) is to study the behaviour of PWR-type fuel rods at high burnup, equipped with an 'advanced' cladding, under Reactivity Initiated Accident (RIA) conditions (such as the scenario of a control rod ejection). Within the framework of this programme, piloted and funded by the French Institute of Nuclear Radioprotection andmore » Safety (IRSN), ten tests are to be conducted with a frequency of two tests per year. The LPRE laboratory of the CEA which is in charge of the Preparation, realisation and breakdown of the test results studies the possibility to set up a new test analysis based on the processing of signals coming from sensors placed within the test equipment. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream of the test device. Studies showed the interest to realize temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena as the rod failure or the test shutdown (i.e. scram). The work presented in this article results from the will to consolidate these studies. Since the main phenomenon to be tracked is the fuel rod failure, the aim would be to highlight specific events which would have been precursors of the rod failure in order to use in the future these signals for further interpretation. Such an antecedent works resumption leads to a better understanding of the experimental needs and constitutes a good initial state to prepare the new very fast digital data acquisition systems. Although all the raw data are accessible in the form of text files, analyses and graphics representations were not straightforward to reproduce from the ancient studies since that, on one hand, people who were in charge of the original work left the laboratory and on the other hand because it is not easy when the time passes, even with our own work, to be able to remember the steps of data manipulations and the exact setup: - During the ancient experiments the use of analog data acquisition systems required to digitize tapes to be able to realize computer treatments. That had had for consequence to lose the initial dating. This one must be correctly edited to do temporal comparisons. - Analyses require functions for calculations whose parameters has to be well-known to reach the same results. We thus wished to manage our workflow in the idea that it can be easily reproducible on all the experiments. The object of the work presented in this article was to put in practice this strong bind between the data, treatments and generation of the document in order not to hesitate to do the iteration principle in action. We do not have to be afraid by the data driven analyses. According to the philosophy of the literate programming, the text of the technical document is woven with the computer code that produces all the printed output as tables, graphs for the study eliminating hence the unrealistic cut and paste. This difficulty is not specific to the nuclear domain. For many years, researchers have been worked out solutions to this mundane issue. And, presently, new technologies and high-level programming languages offer us actual answers. We will firstly present the tools applied in our laboratory to implement this workflow, then we will describe the global perception carried out to continue the study of the Acoustic Emission signals recorded by the two microphones during the fourteen last CABRI R.I.A. test.« 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.

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

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

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

Failure at Frame-Stringer Intersections in PRSEUS Panels

NASA Technical Reports Server (NTRS)

Jegley, Dawn C.

2012-01-01

NASA, the Air Force Research Laboratory and The Boeing Company have worked to develop new low-cost, light-weight composite structures for aircraft. A Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) concept has been developed which offers advantages over traditional metallic structures. In this concept a stitched carbon-epoxy material system has been developed with the potential for reducing the weight and cost of transport aircraft structure by eliminating fasteners, thereby reducing part count and labor. By adding unidirectional carbon rods to the top of stiffeners, the panel becomes more structurally efficient. This combination produces a more damage tolerant design. This study focuses on the intersection between the rod-stiffener and the foam-filled frame in a PRSEUS specimen. Compression loading is considered, which induces stress concentrations at the intersection point that can lead to failures. An experiment with accompanying analysis for a single-frame specimen is described, followed by a parametric study of simple reinforcements to reduce strains in the intersection region.

Evaluation of the Behavior of Technova Corporation Rod-Stiffened Stitched Compression Specimens

NASA Technical Reports Server (NTRS)

Jegley, Dawn C.

2013-01-01

Under Space Act Agreement 1347 between NASA and Technova Corporation, Technova designed and fabricated two carbon-epoxy crippling specimens and NASA loaded them to failure in axial compression. Each specimen contained a pultruded rod stiffener which was held to the specimen skin with through-the-thickness stitches. One of these specimens was designed to be nominally the same as pultruded rod stitched specimens fabricated by Boeing under previous programs. In the other specimen, the rod was prestressed in a Technova manufacturing process to increase its ability to carrying compressive loading. Experimental results demonstrated that the specimen without prestressing carried approximately the same load as the similar Boeing specimens and that the specimen with prestressing carried significantly more load than the specimen without prestressing.

Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

2014-05-01

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are:

Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

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

Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Momore » fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and interaction layer between U–Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.« less

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

FLUID MODERATED REACTOR

DOEpatents

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

1957-10-22

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

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.

Alloyed coatings for dispersion strengthened alloys

NASA Technical Reports Server (NTRS)

Wermuth, F. R.; Stetson, A. R.

1971-01-01

Processing techniques were developed for applying several diffusion barriers to TD-Ni and TD-NiCr. Barrier coated specimens of both substrates were clad with Ni-Cr-Al and Fe-Cr-Al alloys and diffusion annealed in argon. Measurement of the aluminum distribution after annealing showed that, of the readily applicable diffusion barriers, a slurry applied tungsten barrier most effectively inhibited the diffusion of aluminum from the Ni-Cr-Al clad into the TD-alloy substrates. No barrier effectively limited interdiffusion of the Fe-Cr-Al clad with the substrates. A duplex process was then developed for applying Ni-Cr-Al coating compositions to the tungsten barrier coated substrates. A Ni-(16 to 32)Cr-3Si modifier was applied by slurry spraying and firing in vacuum, and was then aluminized by a fusion slurry process. Cyclic oxidation tests at 2300 F resulted in early coating failure due to inadequate edge coverage and areas of coating porosity. EMP analysis showed that oxidation had consumed 70 to 80 percent of the aluminum in the coating in less than 50 hours.

Patient-specific Distraction Regimen to Avoid Growth-rod Failure.

PubMed

Agarwal, Aakash; Jayaswal, Arvind; Goel, Vijay K; Agarwal, Anand K

2018-02-15

A finite element study to establish the relationship between patient's curve flexibility (determined using curve correction under gravity) in juvenile idiopathic scoliosis and the required distraction frequency to avoid growth rod fracture, as a function of time. To perform a parametric analysis using a juvenile scoliotic spine model (single mid-thoracic curve with the apex at the eighth thoracic vertebra) and establish the relationship between curve flexibility (determined using curve correction under gravity) and the distraction interval that allows a higher factor of safety for the growth rods. Previous studies have shown that frequent distraction with smaller magnitude of distractions are less likely to result in rod failure. However there has not been any methodology or a chart provided to apply this knowledge on to the individual patients that undergo the treatment. This study aims to fill in that gap. The parametric study was performed by varying the material properties of the disc, hence altering the axial stiffness of the scoliotic spine model. The stresses on the rod were found to increase with increased axial stiffness of the spine, and this resulted in the increase of required optimal frequency to achieve a factor of safety of two for growth rods. A relationship between the percentage correction in Cobb's angle due to gravity alone, and the required distraction interval for limiting the maximum von Mises stress to 255 MPa on the growth rods was established. The distraction interval required to limit the stresses to the selected nominal value reduces with increase in stiffness of the spine. Furthermore, the appropriate distraction interval reduces for each model as the spine becomes stiffer with time (autofusion). This points to the fact the optimal distraction frequency is a time-dependent variable that must be achieved to keep the maximum von Mises stress under the specified factor of safety. The current study demonstrates the possibility of translating fundamental information from finite element modeling to the clinical arena, for mitigating the occurrence of growth rod fracture, that is, establishing a relationship between optimal distraction interval and curve flexibility (determined using curve correction under gravity). N/A.

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

LOGIC CIRCUIT

DOEpatents

Strong, G.H.; Faught, M.L.

1963-12-24

A device for safety rod counting in a nuclear reactor is described. A Wheatstone bridge circuit is adapted to prevent de-energizing the hopper coils of a ball backup system if safety rods, sufficient in total control effect, properly enter the reactor core to effect shut down. A plurality of resistances form one arm of the bridge, each resistance being associated with a particular safety rod and weighted in value according to the control effect of the particular safety rod. Switching means are used to switch each of the resistances in and out of the bridge circuit responsive to the presence of a particular safety rod in its effective position in the reactor core and responsive to the attainment of a predetermined velocity by a particular safety rod enroute to its effective position. The bridge is unbalanced in one direction during normal reactor operation prior to the generation of a scram signal and the switching means and resistances are adapted to unbalance the bridge in the opposite direction if the safety rods produce a predetermined amount of control effect in response to the scram signal. The bridge unbalance reversal is then utilized to prevent the actuation of the ball backup system, or, conversely, a failure of the safety rods to produce the predetermined effect produces no unbalance reversal and the ball backup system is actuated. (AEC)

Design and evaluation of a bolted joint for a discrete carbon-epoxy rod-reinforced hat section

NASA Technical Reports Server (NTRS)

Rousseau, Carl Q.; Baker, Donald J.

1996-01-01

The use of prefabricated pultruded carbon-epoxy rods has reduced the manufacturing complexity and costs of stiffened composite panels while increasing the damage tolerance of the panels. However, repairability of these highly efficient discrete stiffeners has been a concern. Design, analysis, and test results are presented in this paper for a bolted-joint repair for the pultruded rod concept that is capable of efficiently transferring axial loads in a hat-section stiffener on the upper skin segment of a heavily loaded aircraft wing component. A tension and a compression joint design were evaluated. The tension joint design achieved approximately 1.0% strain in the carbon-epoxy rod-reinforced hat-section and failed in a metal fitting at 166% of the design ultimate load. The compression joint design failed in the carbon-epoxy rod-reinforced hat-section test specimen area at approximately 0.7% strain and at 110% of the design ultimate load. This strain level of 0.7% in compression is similar to the failure strain observed in previously reported carbon-epoxy rod-reinforced hat-section column tests.

Ultrasonic Nondestructive Evaluation of Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) During Large-Scale Load Testing and Rod Push-Out Testing

NASA Technical Reports Server (NTRS)

Johnston, Patrick H.; Juarez, Peter D.

2016-01-01

The Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) is a structural concept developed by the Boeing Company to address the complex structural design aspects associated with a pressurized hybrid wing body (HWB) aircraft configuration. The HWB has long been a focus of NASA's environmentally responsible aviation (ERA) project, following a building block approach to structures development, culminating with the testing of a nearly full-scale multi-bay box (MBB), representing a segment of the pressurized, non-circular fuselage portion of the HWB. PRSEUS is an integral structural concept wherein skins, frames, stringers and tear straps made of variable number of layers of dry warp-knit carbon-fiber stacks are stitched together, then resin-infused and cured in an out-of-autoclave process. The PRSEUS concept has the potential for reducing the weight and cost and increasing the structural efficiency of transport aircraft structures. A key feature of PRSEUS is the damage-arresting nature of the stitches, which enables the use of fail-safe design principles. During the load testing of the MBB, ultrasonic nondestructive evaluation (NDE) was used to monitor several sites of intentional barely-visible impact damage (BVID) as well as to survey the areas surrounding the failure cracks after final loading to catastrophic failure. The damage-arresting ability of PRSEUS was confirmed by the results of NDE. In parallel with the large-scale structural testing of the MBB, mechanical tests were conducted of the PRSEUS rod-to-overwrap bonds, as measured by pushing the rod axially from a short length of stringer.

77 FR 66409 - Airworthiness Directives; Diamond Aircraft Industries GmbH Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-05

... of an unsuitable self-locking nut on the bell crank of the elevator push rod that can cause failure... in the tail section of the fuselage, as a result of installation of a non-suitable self-locking nut... identified that its current configuration has a failure potential when components such as thin self-securing...

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

The Lymphatic Phenotype of Lung Allografts in Patients With Bronchiolitis Obliterans Syndrome and Restrictive Allograft Syndrome.

PubMed

Traxler, Denise; Schweiger, Thomas; Schwarz, Stefan; Schuster, Magdalena Maria; Jaksch, Peter; Lang, Gyoergy; Birner, Peter; Klepetko, Walter; Ankersmit, Hendrik Jan; Hoetzenecker, Konrad

2017-02-01

Chronic lung allograft dysfunction (CLAD), presenting as bronchiolitis obliterans syndrome (BOS) or restrictive allograft syndrome (RAS) is the major limiting factor of long-term survival in lung transplantation. Its pathogenesis is still obscure. In BOS, persistent alloimmune injury and chronic airway inflammation are suggested. One of the main tasks of the lymphatic vessel (LV) system is the promotion of immune cell trafficking. The formation of new LVs has been shown to trigger chronic allograft rejection in kidney transplants. We therefore sought to address the role of lymphangiogenesis in CLAD. Formalin-fixed paraffin-embedded tissue samples of 22 patients receiving a lung retransplantation due to BOS or RAS were collected. Lymphatic vessel density (LVD) was determined by immunohistochemical staining for podoplanin. Lung tissue obtained from 13 non-CLAD patients served as control. The impact of LVD on graft survival was assessed. Lymphatic vessel density in CLAD patients did not differ from those in control subjects (median number of LVs per bronchiole: 4.75 (BOS), 6.47 (RAS), 4.25 (control), P = 0.97). Moreover, the number of LVs was not associated with regions of cellular infiltrates (median number of LVs per bronchiole: with infiltrates, 5.00 (BOS), 9.00 (RAS), 4.00 (control), P = 0.62; without infiltrates, 4.5 (BOS), 0.00 (RAS), 4.56 (control), P = 0.74). Lymphatic vessel density did not impact the time to development of BOS or RAS in lung transplantation (low vs high LVD: 38.5 vs 86.0 months, P = 0.15 [BOS]; 60.5 vs 69.5 months, P = 0.80 [RAS]). Unlike chronic organ failure in kidney transplantation, lymphangiogenesis is not altered in CLAD patients. Our findings highlight unique immunological processes leading to BOS and RAS.

Displaced electrode process for welding

DOEpatents

Heichel, L.J.

1975-08-26

A method is described for the butt-welding of a relatively heavy mass to a relatively small mass such as a thin-wall tube. In butt-welding heat is normally applied at the joint between the two pieces which are butt-welded together. The application of heat at the joint results in overheating the tube which causes thinning of the tube walls and porosity in the tube material. This is eliminated by displacing the welding electrode away from the seam toward the heavier mass so that heat is applied to the heavy mass and not at the butt seam. Examples of the parameters used in welding fuel rods are given. The cladding and end plugs were made of Zircalloy. The electrode used was of 2 percent thoriated tungsten. (auth)

Analysis and Test of Repair Concepts for a Carbon-Rod Reinforced Laminate

NASA Technical Reports Server (NTRS)

Baker, Donald J.; Rousseau, Carl Q.

2000-01-01

The use of pultruded carbon-epoxy rods for the reinforcement of composite laminates in some structures results in an efficient structural concept. The results of an analytical and experimental investigation of repair concepts of completely severed carbon-epoxy rods is presented. Three repair concepts are considered: (a) bonded repair with outside moldline and inside moldline doublers; (b) bonded repair with fasteners, and (c) bonded repair with outside moldline doubler only. The stiffness of the repairs was matched with the stiffness of the baseline specimen. The failure strains for the bonded repair with fasteners and the bonded repair with an outside moldline doubler exceeded a target design strain set for the repair concepts.

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.

NUT SCREW MECHANISMS

DOEpatents

Glass, J.A.F.

1958-07-01

A reactor control mechanism is described wherein the control is achieved by the partial or total withdrawal of the fissile material which is in the form of a fuel rod. The fuel rod is designed to be raised and lowered from the reactor core area by means of two concentric ball nut and screw assemblies that may telescope one within the other. These screw mechanisms are connected through a magnetic clutch to a speed reduction gear and an accurately controllable prime motive source. With the clutch energized, the fuel rod may be moved into the reactor core area, and fine adjustments may be made through the reduction gearing. However, in the event of a power failure or an emergency signal, the magnetic clutch will become deenergized, and the fuel rod will drop out of the core area by the force of gravity, thus shutting down the operation of the reactor.

Safety control circuit for a neutronic reactor

DOEpatents

Ellsworth, Howard C.

2004-04-27

A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

Blister Threshold Based Thermal Limits for the U-Mo Monolithic Fuel System

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

D. M. Wachs; I. Glagolenko; F. J. Rice

2012-10-01

Fuel failure is most commonly induced in research and test reactor fuel elements by exposure to an under-cooled or over-power condition that results in the fuel temperature exceeding a critical threshold above which blisters form on the plate. These conditions can be triggered by normal operational transients (i.e. temperature overshoots that may occur during reactor startup or power shifts) or mild upset events (e.g., pump coastdown, small blockages, mis-loading of fuel elements into higher-than-planned power positions, etc.). The rise in temperature has a number of general impacts on the state of a fuel plate that include, for example, stress relaxationmore » in the cladding (due to differential thermal expansion), softening of the cladding, increased mobility of fission gases, and increased fission-gas pressure in pores, all of which can encourage the formation of blisters on the fuel-plate surface. These blisters consist of raised regions on the surface of fuel plates that occur when the cladding plastically deforms in response to fission-gas pressure in large pores in the fuel meat and/or mechanical buckling of the cladding over damaged regions in the fuel meat. The blister temperature threshold decreases with irradiation because the mechanical properties of the fuel plate degrade while under irradiation (due to irradiation damage and fission-product accumulation) and because the fission-gas inventory progressively increases (and, thus, so does the gas pressure in pores).« less

Design and Evaluation of a Bolted Joint for a Discrete Carbon-Epoxy Rod-Reinforced Hat Section

NASA Technical Reports Server (NTRS)

Baker, Donald J.; Rousseau, Carl Q.

1996-01-01

The use of pre-fabricated pultruded carbon-epoxy rods has reduced the manufacturing complexity and costs of stiffened composite panels while increasing the damage tolerance of the panels. However, repairability of these highly efficient discrete stiffeners has been a concern. Design, analysis, and test results are presented in this paper for a bolted-joint repair for the pultruded rod concept that is capable of efficiently transferring axial loads in a hat-section stiffener on the upper skin segment of a heavily loaded aircraft wing component. A tension and a compression joint design were evaluated. The tension joint design achieved approximately 1.0 percent strain in the carbon-epoxy rod-reinforced hat-section and failed in a metal fitting at 166 percent of the design ultimate load. The compression joint design failed in the carbon-epoxy rod-reinforced hat-section test specimen area at approximately 0.7 percent strain and at 110 percent of the design ultimate load. This strain level of 0.7 percent in compression is similar to the failure strain observed in previously reported carbon-epoxy rod-reinforced hat-section column tests.

Failure analysis of energy storage spring in automobile composite brake chamber

NASA Astrophysics Data System (ADS)

Luo, Zai; Wei, Qing; Hu, Xiaofeng

2015-02-01

This paper set energy storage spring of parking brake cavity, part of automobile composite brake chamber, as the research object. And constructed the fault tree model of energy storage spring which caused parking brake failure based on the fault tree analysis method. Next, the parking brake failure model of energy storage spring was established by analyzing the working principle of composite brake chamber. Finally, the data of working load and the push rod stroke measured by comprehensive test-bed valve was used to validate the failure model above. The experimental result shows that the failure model can distinguish whether the energy storage spring is faulted.

Surface modification techniques for increased corrosion tolerance of zirconium fuel cladding

NASA Astrophysics Data System (ADS)

Carr, James Patrick, IV

Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively employed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neutron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objectives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits. This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and considering materials and processes for modifying the surface of zircaloy fuel cladding. This innovative research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely. Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their microstructural and mechanical properties. To test their application for use in corrosive atmospheres, the corrosion behaviors are also compared in steam, water, and boric-acid environments. Various methods of surface modification were attempted in this investigation, including dip coating, diffusion bonding, casting, sputtering, and evaporation. The benefits and drawbacks of each method are discussed with respect to manufacturing and economic limits. Characterization techniques utilized in this work include optical microscopy, scanning electron microscopy, energy-dispersive spectroscopy, X-ray diffraction, nanoindentation, adhesion testing, and atomic force microscopy. The composition, microstructure, hardness, modulus, and coating adhesion were studied to provide encompassing properties to determine suitable comparisons and to choose an ideal method to scale to industrial applications. The experiments, results, and detailed discussions are presented in the following chapters of this dissertation research.

Multiscale Modeling of Fracture in an SiO2 Nanorod

NASA Astrophysics Data System (ADS)

Mallik, Aditi

2005-11-01

The fracture of a 108 particle SiO2 nanorod under uniaxial strain is described using an NDDO quantum mechanics. The stress -- strain curve to failure is calculated as a function of strain rate to show a domain that is independent of strain rate. A pair potential for use in classical MD is constructed such that the elastic portion of the quantum curve is reproduced. However, it is shown that the classical analysis does not describe accurately the large strain behavior and failure. Finally, a composite rod is constructed with a small subsystem described by quantum mechanics and the remainder described by classical MD ^1. The stress -- strain curves for the classical, quantum, and composite rods are compared and contrasted. 1. ``Multiscale Modeling of Materials -- Concepts and Illustration'', A. Mallik, K. Runge, J. Dufty, and H-P Cheng, cond-mat 0507558.

Posterior Double Vertebral Column Resections Combined with Satellite Rod Technique to Correct Severe Congenital Angular Kyphosis.

PubMed

Sun, Xu; Zhu, Ze-Zhang; Chen, Xi; Liu, Zhen; Wang, Bin; Qiu, Yong

2016-08-01

This paper presents a highly challenging technique involving posterior double vertebral column resections (VCRs) and satellite rods placement. This was a young adult case with severe angular thoracolumbar kyphosis of 101 degrees, secondary to anterior segmentation failure from T11 to L1 . There were hemivertebrae at T11 and T12 , and a wedged vertebra at L1 . He received double VCRs at T12 and T11 and instrumented fusion from T6 to L4 via a posterior only approach. Autologous grafts and a cage were placed between the bony surfaces of the osteotomy gap. Once closure of osteotomy was achieved, bilateral permanent CoCr rods were placed with addition of satellite rods. Postoperative X-ray demonstrated marked correction of kyphosis. On the 10(th) days after surgery, the patient was able to walk without assistance. In conclusion, double VCRs are effective to correct severe angular kyphosis, and addition of satellite rods may be imperative to enhance instrumentation strength and thus prevent correction loss. © 2016 Chinese Orthopaedic Association and John Wiley & Sons Australia, Ltd.

Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

NASA Astrophysics Data System (ADS)

Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

2013-09-01

Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

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

Gas gun driven dynamic fracture and fragmentation of Ti-6Al-4V cylinders

NASA Astrophysics Data System (ADS)

Jones, D. R.; Chapman, D. J.; Eakins, D. E.

2014-05-01

The dynamic fracture and fragmentation of a material is a complex late stage phenomenon occurring in many shock loading scenarios. Improving our predictive capability depends upon exercising our current failure models against new loading schemes and data. We present axially-symmetric high strain rate (104 s-1) expansion of Ti-6Al-4V cylinders using a single stage light gas gun technique. A steel ogive insert was located inside the target cylinder, into which a polycarbonate rod was launched. Deformation of this rod around the insert drives the cylinder into rapid expansion. This technique we have developed facilitates repeatable loading, independent of the temperature of the sample cylinder, with straightforward adjustment of the radial strain rate. Expansion velocity was measured with multiple channels of photon Doppler velocimetry. High speed imaging was used to track the overall expansion process and record strain to failure and crack growth. Results from a cylinder at a temperature of 150 K are compared with work at room temperature, examining the deformation, failure mechanisms and differences in fragmentation.

Comparison of thermal compatibility between atomized and comminuted U{sub 3}Si dispersion fuels

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

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu

1997-08-01

Thermal compatibility of atomized U{sub 3}Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500{degrees}C, and compared with that of comminuted U{sub 3}Si. Atomized U{sub 3}Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U{sub 3}Si and Al occurred along the grain boundaries and deformation bands in U{sub 3}Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, andmore » reduced the volume expansion.« less

Fretting corrosion behavior of nitinol spinal rods in conjunction with titanium pedicle screws.

PubMed

Lukina, Elena; Kollerov, Mikhail; Meswania, Jay; Khon, Alla; Panin, Pavel; Blunn, Gordon W

2017-03-01

Untypical corrosion damage including erosions combined with the build-up of titanium oxide as a corrosion product on the surface of explanted Nitinol spinal rods in the areas where it was in contact with titanium pedicle screw head is reported. It was suggested that Nitinol rods might have inferior fretting corrosion resistance compared with that made of titanium or CoCr. Fretting corrosion of Nitinol spinal rods with titanium (Ti6Al4V) pedicle screws were tested in-vitro by conducting a series of potentiostatic measurements of the peak-to-peak values of fretting corrosion current under bending in a 10% solution of calf serum in PBS. The test included Nitinol rods locked in titanium pedicle screws of different designs. Performance of commercially available titanium (Ti6Al4V) and CoCr spinal rods was also investigated for a comparison. Corrosion damage observed after the in-vitro tests was studied using SEM and EDAX analysis and was compared with patterns on Nitinol rods retrieved 12months after initial surgery. Metal ions level was measured in the test media after in-vitro experiments and in the blood and tissues of the patients who had the rods explanted. The results of this study revealed that Nitinol spinal rods locked in Ti pedicle screws are susceptible to fretting corrosion demonstrating higher fretting corrosion current compared with commercially used Ti6Al4V and CoCr rods. On the surface of Nitinol rods after in-vitro tests and on those retrieved from the patients similar corrosion patterns were observed. Improved resistance to fretting corrosion was observed with Nitinol rods in the in-vitro tests where pedicle screws were used with a stiffer locking mechanism. Since the development of the localized corrosion damage might increase the risk of premature fatigue failure of the rods and result in leaching of Ni ions, it is concluded that Nitinol rods should not be used in conjunction with Ti pedicle screws without special protection especially where the design provides a high degree of mobility to the rods. Copyright © 2016 Elsevier B.V. All rights reserved.

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

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

McKinnon, M.A.; Doherty, A.L.

1997-06-01

This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release frommore » the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs.« less

Reduced yield stress for zirconium exposed to iodine: Reactive force field simulation

DOE PAGES

Rossi, Matthew L.; Taylor, Christopher D.; van Duin, Adri C. T.

2014-11-04

Iodine-induced stress-corrosion cracking (ISCC), a known failure mode for nuclear fuel cladding, occurs when iodine generated during the irradiation of a nuclear fuel pellet escapes the pellet through diffusion or thermal cracking and chemically interacts with the inner surface of the clad material, inducing a subsequent effect on the cladding’s resistance to mechanical stress. To complement experimental investigations of ISCC, a reactive force field (ReaxFF) compatible with the Zr-I chemical and materials systems has been developed and applied to simulate the impact of iodine exposure on the mechanical strength of the material. The study shows that the material’s resistance tomore » stress (as captured by the yield stress of a high-energy grain boundary) is related to the surface coverage of iodine, with the implication that ISCC is the result of adsorption-enhanced decohesion.« less

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2013 CFR

2013-01-01

... elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design..., and functions of reactivity control mechanisms and instrumentation. (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and...

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2014 CFR

2014-01-01

... elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design..., and functions of reactivity control mechanisms and instrumentation. (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and...

Erratic tacrolimus exposure, assessed using the standard deviation of trough blood levels, predicts chronic lung allograft dysfunction and survival.

PubMed

Gallagher, Harry M; Sarwar, Ghulam; Tse, Tracy; Sladden, Timothy M; Hii, Esmond; Yerkovich, Stephanie T; Hopkins, Peter M; Chambers, Daniel C

2015-11-01

Erratic tacrolimus blood levels are associated with liver and kidney graft failure. We hypothesized that erratic tacrolimus exposure would similarly compromise lung transplant outcomes. This study assessed the effect of tacrolimus mean and standard deviation (SD) levels on the risk of chronic lung allograft dysfunction (CLAD) and death after lung transplantation. We retrospectively reviewed 110 lung transplant recipients who received tacrolimus-based immunosuppression. Cox proportional hazard modeling was used to investigate the effect of tacrolimus mean and SD levels on survival and CLAD. At census, 48 patients (44%) had developed CLAD and 37 (34%) had died. Tacrolimus SD was highest for the first 6 post-transplant months (median, 4.01; interquartile range [IQR], 3.04-4.98 months) before stabilizing at 2.84 μg/liter (IQR, 2.16-4.13 μg/liter) between 6 and 12 months. The SD then remained the same (median, 2.85; IQR, 2.00-3.77 μg/liter) between 12 and 24 months. A high mean tacrolimus level 6 to 12 months post-transplant independently reduced the risk of CLAD (hazard ratio [HR], 0.74; 95% confidence interval [CI], 0.63-0.86; p < 0.001) but not death (HR, 0.96; 95% CI, 0.83-1.12; p = 0.65). In contrast, a high tacrolimus SD between 6 and 12 months independently increased the risk of CLAD (HR, 1.46; 95% CI, 1.23-1.73; p < 0.001) and death (HR, 1.27; 95% CI, 1.08-1.51; p = 0.005). Erratic tacrolimus levels are a risk factor for poor lung transplant outcomes. Identifying and modifying factors that contribute to this variability may significantly improve outcomes. Copyright © 2015 International Society for Heart and Lung Transplantation. Published by Elsevier Inc. All rights reserved.

Hanging core support system for a nuclear reactor

DOEpatents

Burelbach, James P.; Kann, William J.; Pan, Yen-Cheng; Saiveau, James G.; Seidensticker, Ralph W.

1987-01-01

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

Optical coherence tomography for nondestructive evaluation of fuel rod degradation

NASA Astrophysics Data System (ADS)

Renshaw, Jeremy B.; Jenkins, Thomas P.; Buckner, Benjamin D.; Friend, Brian

2015-03-01

Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal.

Composition-Property Correlation in B2O3-SiO2 Preform Rods Produced Using Modified Chemical Vapor Deposition Technique

NASA Astrophysics Data System (ADS)

Islam, Mohammad; Saleem, Muhammad Rizwan

2012-02-01

Due to unique optical properties of high birefringent (Hi-Bi) fibers for sensing and coherent optical communications, there is a strong interest in process optimization at preform fabrication and fiber drawing stages. Boron-doped silica cladding acts as stress-applying part resulting in polarization properties of Hi-Bi fibers that are strongly dependent on chemical composition. Using modified chemical vapor deposition (MCVD) technique, B2O3-doped silica preform rods were synthesized under different precursor gas flow conditions. Qualitative information about B2O3-SiO2 system composition was derived from etching behavior in nonbuffered HF solution and subsequent microstructural examination using scanning electron microscope. Significant degree of B2O3 incorporation was seen in case of high BCl3:SiCl4 ratio and mild oxygen-deficient processing conditions. Increasing the B2O3 content to ~26 mol% led to a corresponding increase in coefficient of thermal expansion (CTE) to a maximum value of 2.35 ppm/K. The value of refractive index (RI), on the other hand, was found to decrease with increased B2O3 incorporation. A qualitative correlation between B2O3 and SiO2 system composition and physical properties such as CTE and RI was established.

Analytical Modeling Tool for Design of Hydrocarbon Sensitive Optical Fibers

PubMed Central

Vahdati, Nader; Lawand, Lydia

2017-01-01

Pipelines are the main transportation means for oil and gas products across large distances. Due to the severe conditions they operate in, they are regularly inspected using conventional Pipeline Inspection Gages (PIGs) for corrosion damage. The motivation for researching a real-time distributed monitoring solution arose to mitigate costs and provide a proactive indication of potential failures. Fiber optic sensors with polymer claddings provide a means of detecting contact with hydrocarbons. By coating the fibers with a layer of metal similar in composition to that of the parent pipeline, corrosion of this coating may be detected when the polymer cladding underneath is exposed to the surrounding hydrocarbons contained within the pipeline. A Refractive Index (RI) change occurs in the polymer cladding causing a loss in intensity of a traveling light pulse due to a reduction in the fiber’s modal capacity. Intensity losses may be detected using Optical Time Domain Reflectometry (OTDR) while pinpointing the spatial location of the contact via time delay calculations of the back-scattered pulses. This work presents a theoretical model for the above sensing solution to provide a design tool for the fiber optic cable in the context of hydrocarbon sensing following corrosion of an external metal coating. Results are verified against the experimental data published in the literature. PMID:28956847

Analytical Modeling Tool for Design of Hydrocarbon Sensitive Optical Fibers.

PubMed

Al Handawi, Khalil; Vahdati, Nader; Shiryayev, Oleg; Lawand, Lydia

2017-09-28

Pipelines are the main transportation means for oil and gas products across large distances. Due to the severe conditions they operate in, they are regularly inspected using conventional Pipeline Inspection Gages (PIGs) for corrosion damage. The motivation for researching a real-time distributed monitoring solution arose to mitigate costs and provide a proactive indication of potential failures. Fiber optic sensors with polymer claddings provide a means of detecting contact with hydrocarbons. By coating the fibers with a layer of metal similar in composition to that of the parent pipeline, corrosion of this coating may be detected when the polymer cladding underneath is exposed to the surrounding hydrocarbons contained within the pipeline. A Refractive Index (RI) change occurs in the polymer cladding causing a loss in intensity of a traveling light pulse due to a reduction in the fiber's modal capacity. Intensity losses may be detected using Optical Time Domain Reflectometry (OTDR) while pinpointing the spatial location of the contact via time delay calculations of the back-scattered pulses. This work presents a theoretical model for the above sensing solution to provide a design tool for the fiber optic cable in the context of hydrocarbon sensing following corrosion of an external metal coating. Results are verified against the experimental data published in the literature.

Late Complication of Surgically Treated Atlantoaxial Instability: Occipital Bone Erosion Induced by Protruded Fixed Titanium Rod: A Case Report

PubMed Central

Nakao, Yaoki; Shimokawa, Nobuyuki; Morisako, Hiroki; Tsukazaki, Yuji; Terada, Aiko; Nakajo, Kosuke; Fu, Yoshihiko

2014-01-01

Objective Polyaxial screw-rod fixation of C1-C2 is a relatively new technique to treat atlantoaxial instability, and there have been few reports in the literature outlining all possible complications. The purpose of this case report is to present the occurrence and management of occipital bone erosion induced by the protruded rostral part of a posterior atlantoaxial screw-rod construct causing headache. Clinical Features A 70-year-old Asian man with rheumatoid arthritis initially presented to our institution with atlantoaxial instability causing progressive quadraparesis and neck pain. Intervention and Outcome Posterior atlantoaxial instrumented fixation using C1 lateral mass screws in conjunction with C2 pedicle screws was performed to stabilize these segments. Postoperatively, the patient regained the ability to independently walk and had no radiographic evidence of instrumentation hardware failure and excellent sagittal alignment. However, despite a well-stabilized fusion, the patient began to complain of headache during neck extension. Follow-up imaging studies revealed left occipital bone erosion induced by a protruded titanium rod fixed with setscrews. During revision surgery, the rod protrusion was modified and the headaches diminished. Conclusion This case demonstrates that occipital bone erosion after posterior atlantoaxial fixation causing headache may occur. The principal cause of bone erosion in this case was rod protrusion. Although posterior atlantoaxial fixation using the screw-rod system was selected to manage atlantoaxial instability because it has less complications than other procedures, surgeons should pay attention that the length of the rod protrusion should not exceed 2 mm. PMID:25435842

Determination of Graphite-Liquid-Vapor Triple Point by Laser Heating

DTIC Science & Technology

1976-01-30

difficulties in the temperature measure- ments, which were made with an optical pyrometer . He considered that the failure of graphite rod was caused by...temperature measurements were made with a calibrated optical pyrometer . Spherical shiny frozen droplets of graphite, 1 to 3 mm in diameter, indicated...0.8 mm in diameter and 10 mm long in argon until failure by rupture occurred. They measured the tempera- ture with a two-color pyrometer . The

Cost analysis of magnetically controlled growing rods compared with traditional growing rods for early-onset scoliosis in the US: an integrated health care delivery system perspective

PubMed Central

Polly, David W; Ackerman, Stacey J; Schneider, Karen; Pawelek, Jeff B; Akbarnia, Behrooz A

2016-01-01

Purpose Traditional growing rod (TGR) for early-onset scoliosis (EOS) is effective but requires repeated invasive surgical lengthenings under general anesthesia. Magnetically controlled growing rod (MCGR) is lengthened noninvasively using a hand-held magnetic external remote controller in a physician office; however, the MCGR implant is expensive, and the cumulative cost savings have not been well studied. We compared direct medical costs of MCGR and TGR for EOS from the US integrated health care delivery system perspective. We hypothesized that over time, the MCGR implant cost will be offset by eliminating repeated TGR surgical lengthenings. Methods For both TGR and MCGR, the economic model estimated the cumulative costs for initial implantation, lengthenings, revisions due to device failure, surgical-site infections, device exchanges (at 3.8 years), and final fusion, over a 6-year episode of care. Model parameters were estimated from published literature, a multicenter EOS database of US institutions, and interviews. Costs were discounted at 3.0% annually and represent 2015 US dollars. Results Of 1,000 simulated patients over 6 years, MCGR was associated with an estimated 270 fewer deep surgical-site infections and 197 fewer revisions due to device failure compared with TGR. MCGR was projected to cost an additional $61 per patient over the 6-year episode of care compared with TGR. Sensitivity analyses indicated that the results were sensitive to changes in the percentage of MCGR dual rod use, months between TGR lengthenings, percentage of hospital inpatient (vs outpatient) TGR lengthenings, and MCGR implant cost. Conclusion Cost neutrality of MCGR to TGR was achieved over the 6-year episode of care by eliminating repeated TGR surgical lengthenings. To our knowledge, this is the first cost analysis comparing MCGR to TGR – from the US provider perspective – which demonstrates the efficient provision of care with MCGR. PMID:27695352

Overview of the U.S. DOE Accident Tolerant Fuel Development Program

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

Jon Carmack; Frank Goldner; Shannon M. Bragg-Sitton

2013-09-01

The United States Fuel Cycle Research and Development Advanced Fuels Campaign has been given the responsibility to conduct research and development on enhanced accident tolerant fuels with the goal of performing a lead test assembly or lead test rod irradiation in a commercial reactor by 2022. The Advanced Fuels Campaign has defined fuels with enhanced accident tolerance as those that, in comparison with the standard UO2-Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining ormore » improving the fuel performance during normal operations and operational transients, as well as design-basis and beyond design-basis events. This paper provides an overview of the FCRD Accident Tolerant Fuel program. The ATF attributes will be presented and discussed. Attributes identified as potentially important to enhance accident tolerance include reduced hydrogen generation (resulting from cladding oxidation), enhanced fission product retention under severe accident conditions, reduced cladding reaction with high-temperature steam, and improved fuel-cladding interaction for enhanced performance under extreme conditions. To demonstrate the enhanced accident tolerance of candidate fuel designs, metrics must be developed and evaluated using a combination of design features for a given LWR design, potential improvements to that design, and the design of an advanced fuel/cladding system. The aforementioned attributes provide qualitative guidance for parameters that will be considered for fuels with enhanced accident tolerance. It may be unnecessary to improve in all attributes and it is likely that some attributes or combination of attributes provide meaningful gains in accident tolerance, while others may provide only marginal benefits. Thus, an initial step in program implementation will be the development of quantitative metrics. A companion paper in these proceedings provides an update on the status of establishing these quantitative metrics for accident tolerant LWR fuel.1 The United States FCRD Advanced Fuels Campaign has embarked on an aggressive schedule for development of enhanced accident tolerant LWR fuels. The goal of developing such a fuel system that can be deployed in the U.S. LWR fleet in the next 10 to 20 years supports the sustainability of clean nuclear power generation in the United States.« less

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.

Pedicle screw-rod fixation: a feasible treatment for dogs with severe degenerative lumbosacral stenosis.

PubMed

Tellegen, Anna R; Willems, Nicole; Tryfonidou, Marianna A; Meij, Björn P

2015-12-07

Degenerative lumbosacral stenosis is a common problem in large breed dogs. For severe degenerative lumbosacral stenosis, conservative treatment is often not effective and surgical intervention remains as the last treatment option. The objective of this retrospective study was to assess the middle to long term outcome of treatment of severe degenerative lumbosacral stenosis with pedicle screw-rod fixation with or without evidence of radiological discospondylitis. Twelve client-owned dogs with severe degenerative lumbosacral stenosis underwent pedicle screw-rod fixation of the lumbosacral junction. During long term follow-up, dogs were monitored by clinical evaluation, diagnostic imaging, force plate analysis, and by using questionnaires to owners. Clinical evaluation, force plate data, and responses to questionnaires completed by the owners showed resolution (n = 8) or improvement (n = 4) of clinical signs after pedicle screw-rod fixation in 12 dogs. There were no implant failures, however, no interbody vertebral bone fusion of the lumbosacral junction was observed in the follow-up period. Four dogs developed mild recurrent low back pain that could easily be controlled by pain medication and an altered exercise regime. Pedicle screw-rod fixation offers a surgical treatment option for large breed dogs with severe degenerative lumbosacral stenosis with or without evidence of radiological discospondylitis in which no other treatment is available. Pedicle screw-rod fixation alone does not result in interbody vertebral bone fusion between L7 and S1.

PCI fuel failure analysis: a report on a cooperative program undertaken by Pacific Northwest Laboratory and Chalk River Nuclear Laboratories.

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

Mohr, C.L.; Pankaskie, P.J.; Heasler, P.G.

Reactor fuel failure data sets in the form of initial power (P/sub i/), final power (P/sub f/), transient increase in power (..delta..P), and burnup (Bu) were obtained for pressurized heavy water reactors (PHWRs), boiling water reactors (BWRs), and pressurized water reactors (PWRs). These data sets were evaluated and used as the basis for developing two predictive fuel failure models, a graphical concept called the PCI-OGRAM, and a nonlinear regression based model called PROFIT. The PCI-OGRAM is an extension of the FUELOGRAM developed by AECL. It is based on a critical threshold concept for stress dependent stress corrosion cracking. The PROFITmore » model, developed at Pacific Northwest Laboratory, is the result of applying standard statistical regression methods to the available PCI fuel failure data and an analysis of the environmental and strain rate dependent stress-strain properties of the Zircaloy cladding.« less

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

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

Katoh, Yutai; Wilson, Dane F; Forsberg, Charles W

2007-09-01

The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) compositesmore » are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.« less

Review of Findings for Human Performance Contribution to Risk in Operating Events

DTIC Science & Technology

2002-03-01

and loss of DC power. Key to this event was failure to control setpoints on safety-related equipment and failure to maintain the load tap changer...34 Therefore, "to optimize task execution at the job site, it is important to align organizational processes and values." Effective team skills are an...reactor was blocked and the water level rapidly dropped to the automatic low-level scram setpoint . Human Performance Issues Control rods were fully

Failure analysis and evaluation of a six cylinders crankshaft for marine diesel generator

NASA Astrophysics Data System (ADS)

Khaeroman, Haryadi, Gunawan Dwi; Ismail, R.; Kim, Seon Jin

2017-01-01

This paper discusses the failure of a diesel engine crankshaft of a four stroke 6 cylinders, used in a marine diesel generator. A correct analysis and evaluation of the dimension of the crankshaft are very essential to prevent failure of the crankshaft fracture and cracks. The crankshaft is liable to deformation due to misalignment of the main journals bearings. This article presents the result of crankshaft failure analysis by measuring the mean diameter of the rod journal and the main journal, on the wear, out of roundness, taper, etc. The measurement results must be compared with the acceptable value in the engine specification and manual service and also should follow the American Bureau of Shipping (ABS) guidance notes on propulsion shafting alignment. The measurement results of this study show that the main journal diameter of the third cylinder exhibits an excessive wear, 1.35 % above the permissible lowest rate. It also has a taper for 0.23 mm and out of roundness of 0.13 mm. The diameter of the rod journal indicates excessive wear, 1.06 % higher than the permissible lowest rate, the taper of 0.41 mm and out of roundness of 0.65 mm. The crankshaft warpage or run-out journal, the analysis of the crank web deflection are also evaluated and presented in this paper.

Characterizing the dynamic strength of materials for ballistic applications

NASA Astrophysics Data System (ADS)

Cazamias, James Ulysses

We unambiguously verified the hypothesis that normal penetration in brittle materials may be represented as a bi-modal process. The first mode is governed by fundamental strength properties of the target, while the second mode is governed by the fracture kinetics. We investigated the failure response of glass under impact loading. We observed a drop in the failure wave velocity by a factor of 1/2 after unloading. While not unexpected, this drop had not been clearly observed previously. In contradiction to literature values, we observed a drop in sound speed behind the failure wave. Finally, despite the common perception that the failed material is comminuted, we observed a finite tensile strength. We proposed a new variant of the Taylor test using scaled rods to examine strain rate effects. For armor steel, we observed changes in strength greater than what would be expected from a logarithmic dependence of strength on strain rate although not enough to account for scale effects. For tungsten penetrators, we observed that smaller scale tungsten rods appeared to have more work hardening than the large scale rods which might account for scale effects. We examined the square Taylor impact problem. We showed that the square Taylor test is a new way to study shear localization under compressive-shear loading. We performed the first shock characterization of AlON. We observed that the bar impact experiment appears to differentiate between different thicknesses of ceramic tile in qualitative agreement with subscale and full scale penetration experiments. We present data supporting the lower yield strength estimate of 4.3 GPa for alumina. We performed the first bar impact characterization of AlON.

Lateral mass screw-rod fixation of the cervical spine: a prospective clinical series with 1-year follow-up.

PubMed

Deen, H Gordon; Birch, Barry D; Wharen, Robert E; Reimer, Ronald

2003-01-01

Lateral mass plating has become the technique of choice for posterior cervical fixation. Although these systems are safe and reliable, they can be difficult to use in patients with abnormal cervical anatomy; screw placement can be compromised by the fixed hole spacing of the plate; screw back-out and other forms of implant failure can occur; and extension across the cervicothoracic junction can be problematic. To report a series of patients undergoing posterior cervical stabilization with a polyaxial screw-rod construct and to investigate whether this new system offers any advantages over existing methods of fixation. A prospective study evaluating clinical and radiographic parameters in a consecutive series of patients treated with this technique. There were 21 patients in the study group. The surgical indication was cervical spondylosis in 14, trauma in 2, postsurgical kyphosis in 2 and 1 case each of congenital cervicothoracic stenosis, C7-T1 pseudarthrosis and basilar invagination with brainstem compression. Clinical indicators included age, gender, neurologic status, surgical indication and number of levels stabilized. Note was made of whether laminectomy and concomitant anterior reconstructive surgery were performed. Radiographic indicators included early postoperative computed tomography (CT) scan to check for screw placement and plain radiographs at subsequent visits. The participants in this study underwent posterior cervical stabilization using lateral mass screw-rod fixation. Clinical and radiographic assessment was carried out immediately after surgery, and 3, 6 and 12 months after surgery. One-year follow-up was obtained in all cases. A total of 212 screws were implanted in 21 patients. Fixation was carried out over an average of 5.5 spinal segments (range, 2 to 11). The system was successfully implanted in all patients despite the presence of coronal and sagittal plane deformities and/or lateral mass abnormalities in the majority of cases. This system allowed for screw placement in the occiput, C1 lateral mass, C2 pars, C3-C7 lateral masses and upper thoracic pedicles. Early postoperative CT scanning confirmed satisfactory screw placement in all cases. Three patients experienced transient single-level radiculopathy, for an incidence of 1.4% per screw placed. Two patients developed wound seromas requiring evacuation. There were no infections or other wound healing problems. There were no examples of cord or vertebral artery injury, cerebrospinal fluid leak, screw malposition or back-out, loss of alignment or implant failure. When compared with plating techniques, screw-rod fixation appeared to offer several advantages. First, unlike plates, rods proved to be amenable to multiplanar contouring, which is often needed for deformities associated with cervical spondylosis. Second, lateral mass screw placement was more precise because it was not constrained by the hole spacing of the plate. Third, screw back-out and other types of implant failure were not seen. Fourth, the screw-rod system was more easily extended to the occiput and across the cervicothoracic junction. Fifth, the screw-rod system permitted the application of compression, distraction and reduction forces within the construct, to a greater extent than plate systems. The incidence of postoperative radiculopathy was similar to that seen with plate systems. These data indicate that posterior cervical stabilization with polyaxial screw-rod fixation is a safe, straightforward technique that appears to offer some advantages over existing methods of fixation. Results appear to be durable at 1-year follow-up. Benefits are more significant with longer constructs, especially those extending to the occiput or crossing the cervicothoracic junction.

Crack growth through the thickness of thin-sheet Hydrided Zircaloy-4

NASA Astrophysics Data System (ADS)

Raynaud, Patrick A. C.

In recent years, the limits on fuel burnup have been increased to allow an increase in the amount of energy produced by a nuclear fuel assembly thus reducing waste volume and allowing greater capacity factors. As a result, it is paramount to ensure safety after longer reactor exposure times in the case of design-basis accidents, such as reactivity-initiated accidents (RIA). Previously proposed failure criteria do not directly address the particular cladding failure mechanism during a RIA, in which crack initiation in brittle outer-layers is immediately followed by crack growth through the thickness of the thin-wall tubing. In such a case, the fracture toughness of hydrided thin-wall cladding material must be known for the conditions of through-thickness crack growth in order to predict the failure of high-burnup cladding. The fracture toughness of hydrided Zircaloy-4 in the form of thin-sheet has been examined for the condition of through-thickness crack growth as a function of hydride content and distribution at 25°C, 300°C, and 375°C. To achieve this goal, an experimental procedure was developed in which a linear hydride blister formed across the width of a four-point bend specimen was used to inject a sharp crack that was subsequently extended by fatigue pre-cracking. The electrical potential drop method was used to monitor the crack length during fracture toughness testing, thus allowing for correlation of the load-displacement record with the crack length. Elastic-plastic fracture mechanics were used to interpret the experimental test results in terms of fracture toughness, and J-R crack growth resistance curves were generated. Finite element modeling was performed to adapt the classic theories of fracture mechanics applicable to thick-plate specimens to the case of through-thickness crack growth in thin-sheet materials, and to account for non-uniform crack fronts. Finally, the hydride microstructure was characterized in the vicinity of the crack tip by means of digital image processing, so as to understand the influence of the hydride microstructure on fracture toughness, at the various test temperatures. Crack growth occurred through a microstructure which varied within the thickness of the thin-sheet Zircaloy-4 such that the hydrogen concentration and the radial hydride content decreased with increasing distance from the hydride blister. At 25°C, the fracture toughness was sensitive to the changes in hydride microstructure, such that the toughness KJi decreased from 39 MPa√m to 24 MPa√m with increasing hydrogen content and increasing the fraction of radial hydrides. The hydride particles present in the Zircaloy-4 substrate fractured ahead of the crack tip, and crack growth occurred by linking the crack-tip with the next hydride-induced primary void ahead of it. Unstable crack growth was observed at 25°C prior to any stable crack growth in the specimens where the hydrogen content was the highest. At 375°C as well as in most cases at 300°C, the hydride particles were resistant to cracking and the resistance to crack-growth initiation was very high. As a result, for this bend test procedure, crack extension was solely due to crack-tip blunting instead of crack growth in all tests at 375°C and in most cases at 300°C. The lower bound for fracture toughness at these temperatures, the parameter KJPmax, had values of K JPmax˜54MPa√m at both 300°C and 375°C. For cases where stable crack growth occurred at 300°C, the fracture toughness was K Ji˜58MPa√m and the tearing modulus was twice as high as that at 25°C. It is believed that the failure of hydrided Zircaloy-4 thin-wall cladding can be predicted using fracture mechanics analyses when failure occurs by crack growth. This failure mechanism was observed to occur in all cases at 25°C and in some cases at 300°C. However, at more elevated temperatures, such as 375°C, failure will likely occur by a mechanism other than crack growth, possibly by an imperfection-induced shear instability.

Optical coherence tomography for nondestructive evaluation of fuel rod degradation

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

Renshaw, Jeremy B., E-mail: jrenshaw@epri.com; Jenkins, Thomas P., E-mail: tjenkins@metrolaserinc.com; Buckner, Benjamin D., E-mail: tjenkins@metrolaserinc.com

Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such asmore » Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal.« less

TEFLON BELLOWS PULSE GENERATORS FOR SOLVENT EXTRACTION PULSE COLUMNS

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

McCarthy, P.B.

1954-01-01

A Teflon bellows-type pulse generator is described which consists of two 3-in. nominal diameter Teflon bellows mounted on either end of a flanged spool piece and rigidly connected internally by a push rod so that the two of bellows move in tandem. The ends of the two bellows are closed by means of blind flanges. Tne spool piece is designed for insertion in a 6-in. diameter nozzle on a solvent extraction pulse column. The double bellows arrangement constitutes a safety feature to prevent loss of the column contents in the event of failure of the inner bellows in contact withmore » column solution. Failure of the inner bellows may be detected by a conductivity probe mounted in the air space inside of the double bellows assembly. Reciprocating motion is imcrank arm rigidly connected through a cross head and push rod to the face of the external bellows flange. The push rod is guided by means of linear ball bushings. Frequency variation over a range of 30 to 100 cycles/ min.was obtained by use of a Thymotrol-controlled electric motor to drive the crank arm. Variable stroke adjustment (0 to 1-in. range) was possible by adjustment of linkages on the crank arm. A load compensating spring was founnd desirable to counteract the thrust on the push rod resulting tom the static pressure at the bottom of the solvent extraction column. Without the spring, accelerated wear of the bearing on the crank arm occured. The pulse generator operated uneventfully for 1776 hours (6.61x lO/sup 6/ cycles) at a frequency of 62 cycles/min. and a bellows travel of l-in. (equivalent to a displacement of 1.6 in. in a 3-in. diam. column). (auth)« less

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

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

The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNFmore » are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.« less

Hanging core support system for a nuclear reactor. [LMFBR

DOEpatents

Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

1984-04-26

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

Planar dynamics of large-deformation rods under moving loads

NASA Astrophysics Data System (ADS)

Zhao, X. W.; van der Heijden, G. H. M.

2018-01-01

We formulate the problem of a slender structure (a rod) undergoing large deformation under the action of a moving mass or load motivated by inspection robots crawling along bridge cables or high-voltage power lines. The rod is described by means of geometrically exact Cosserat theory which allows for arbitrary planar flexural, extensional and shear deformations. The equations of motion are discretised using the generalised-α method. The formulation is shown to handle the discontinuities of the problem well. Application of the method to a cable and an arch problem reveals interesting nonlinear phenomena. For the cable problem we find that large deformations have a resonance detuning effect on cable dynamics. The problem also offers a compelling illustration of the Timoshenko paradox. For the arch problem we find a stabilising (delay) effect on the in-plane collapse of the arch, with failure suppressed entirely at sufficiently high speed.

Microstructure and wear resistance of Ti-Cu-N composite coating prepared via laser cladding/laser nitriding technology on Ti-6Al-4V alloy

NASA Astrophysics Data System (ADS)

Yang, Yuling; Cao, Shiyin; Zhang, Shuai; Xu, Chuan; Qin, Gaowu

2017-07-01

Ti-Cu-N coatings with three different Cu contents on Ti-6Al-4V alloy (TC4) were obtained via laser cladding together with laser nitriding (LC/LN) technology. Phase constituents, microstructure, microhardness, and wear resistance of the coatings were investigated. The evolution of the coefficients of friction for the three coatings was measured under dry sliding conditions as a function of the revolutions until the coating failure. The results show that the coatings are mainly composed of TiN, CuTi3 and some TiO6 phases dispersed in the matrix. A good metallurgical bonding between the coating and substrate has been successfully obtained. The prepared Ti-Cu-N composite coatings almost doubly enhance the microhardness of the TC4 alloy and reduce the friction down to 1/4-1/2 of the TC4 alloy, and thus significantly improve the wear resistance. The coefficient of friction depends on the Cu content in the coating.

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

Analysis of failed nuclear plant components

NASA Astrophysics Data System (ADS)

Diercks, D. R.

1993-12-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

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

NASA Technical Reports Server (NTRS)

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

2018-01-01

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

Penetration of tungsten-alloy rods into composite ceramic targets: Experiments and 2-D simulations

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

Rosenberg, Z.; Dekel, E.; Hohler, V.

1998-07-10

A series of terminal ballistics experiments, with scaled tungsten-alloy penetrators, was performed on composite targets consisting of ceramic tiles glued to thick steel backing plates. Tiles of silicon-carbide, aluminum nitride, titanium-dibroide and boron-carbide were 20-80 mm thick, and impact velocity was 1.7 km/s. 2-D numerical simulations, using the PISCES code, were performed in order to simulate these shots. It is shown that a simplified version of the Johnson-Holmquist failure model can account for the penetration depths of the rods but is not enough to capture the effect of lateral release waves on these penetrations.

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

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

2006-06-01

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

Behavior of S.A.P. in the Mercury Catalyzed Nitric Acid Dissolution; COMPORTAMENTO DEL S.A.P. ALL'ATTACCO DI SOLUZIONI DI ACIDO NITRICO E NITRATO MERCURICO

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

Beone, G.

1963-10-01

Plates of S.A.P. (sintered Aluminum Powder) were dissolved under different conditions in a nitric acid solution containing mercuric nitrate as a catalyst. These experiments nim at establishing a head-end dissolution process for S.A.P. cladded uranium oxide fuels. The results of preliminary dissolution experiments on simulated fuel rods are also described. The behavior of S.A.P. in the mercury catalyzed nitric acid dissolutions differs strongly from the behavior of aluminum: reaction rates are very low for S.A.P. and the dissolution time borders on being unacceptable in an industrial process. Settling rates of suspended alumina are however favorable. A tentative head end flowsheetmore » lay out for PRO second core fuel elements is included. (auth)« 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.

Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

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

Su'ud, Zaki, E-mail: szaki@fi.itba.c.id; Sekimoto, H., E-mail: hsekimot@gmail.com

2014-09-30

Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature canmore » be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.« less

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.

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

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...

Preventing distal pullout of posterior spine instrumentation in thoracic hyperkyphosis: a biomechanical analysis.

PubMed

Sun, Edward; Alkalay, Ron; Vader, David; Snyder, Brian D

2009-06-01

An in vitro biomechanical study. Compare the mechanical behavior of 5 different constructs used to terminate dual-rod posterior spinal instrumentation in resisting forward flexion moment. Failure of the distal fixation construct can be a significant problem for patients undergoing surgical treatment for thoracic hyperkyphosis. We hypothesize that augmenting distal pedicle screws with infralaminar hooks or sublaminar cables significantly increases the strength and stiffness of these constructs. Thirty-seven thoracolumbar (T12 to L2) calf spines were implanted with 5 configurations of distal constructs: (1) infralaminar hooks, (2) sublaminar cables, (3) pedicle screws, (4) pedicle screws+infralaminar hooks, and (5) pedicle screws+sublaminar cables. Progressive bending moment was applied to each construct until failure. The mode of failure was noted and the construct's stiffness and failure load determined from the load-displacement curves. Bone density and vertebral dimensions were equivalent among the groups (F=0.1 to 0.9, P>0.05). One-way analysis of covariance (adjusted for differences in density and vertebral dimension) demonstrated that all of the screw-constructs (screw, screw+hook, and screw+cable) exhibited significantly higher stiffness and ultimate failure loads compared with either sublaminar hook or cable alone (P<0.05). The screw+hook constructs (109+/-11 Nm/mm) were significantly stiffer than either screws alone (88+/-17 Nm/mm) or screw+cable (98+/-13 Nm/mm) constructs, P<0.05. Screw+cable construct exhibited significantly higher failure load (1336+/-328 N) compared with screw constructs (1102+/-256 N, P<0.05), whereas not statistically different from the screw+hook construct (1220+/-75 N). The cable and hook constructs failed by laminar fracture, screw construct failed in uniaxial shear (pullout), whereas the screws+(hooks or wires) failed by fracture of caudal vertebral body. Posterior dual rod constructs fixed distally using pedicle screws were stiffer and stronger in resisting forward flexion compared with cables or hooks alone. Augmenting these screws with either infralaminar hooks or sublaminar cables provided additional resistance to failure.

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

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

Renier, J.A.

2002-04-17

Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron.more » Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized water reactor fuel core was chosen for the study, and state-of-the-art neutronic reactor core computer codes were used for analysis. Power distribution, fuel burnup, reactivity due to burnable poisons and other fission products, spectrum shift, core reactivity, moderator void coefficients, as well as other parameters were calculated as a function of time and fuel burnup. The results not only showed advantages of separation of burnable poison isotopes but revealed benefits to be achieved by careful selection of the configuration of even naturally occurring elements used as burnable poisons. The savings in terms of additional days of operation is shown in Figure 1, where the savings is plotted for each of six favorable isotopes in the four configurations. The benefit of isotope separation is most dramatic for dysprosium, but even the time savings in the case of gadolinium is several days. For a modern nuclear plant, one day's worth of electricity is worth about one million dollars, so the resulting savings of only a few days is considerable. It is also apparent that the amount of savings depends upon the configuration of the burnable poison.« less

Stability and failure analysis of steering tie-rod

NASA Astrophysics Data System (ADS)

Jiang, GongFeng; Zhang, YiLiang; Xu, XueDong; Ding, DaWei

2008-11-01

A new car in operation of only 8,000 km, because of malfunction, resulting in lost control and rammed into the edge of the road, and then the basic vehicle scrapped. According to the investigation of the site, it was found that the tie-rod of the car had been broken. For the subjective analysis of the accident and identifying the true causes of rupture of the tierod, a series of studies, from the angle of theory to experiment on the bended broken tie-rod, were conducted. The mechanical model was established; the stability of the defective tie-rod was simulated based on ANSYS software. Meanwhile, the process of the accident was simulated considering the effect of destabilization of different vehicle speed and direction of the impact. Simultaneously, macro graphic test, chemical composition analysis, microstructure analysis and SEM analysis of the fracture were implemented. The results showed that: 1) the toughness of the tie-rod is at a normal level, but there is some previous flaws. One quarter of the fracture surface has been cracked before the accident. However, there is no relationship between the flaw and this incident. The direct cause is the dynamic instability leading to the large deformation of impact loading. 2) The declining safety factor of the tie-rod greatly due to the previous flaws; the result of numerical simulation shows that previous flaw is the vital factor of structure instability, on the basis of the comparison of critical loads of the accident tie-rod and normal. The critical load can decrease by 51.3% when the initial defect increases 19.54% on the cross-sectional area, which meets the Theory of Koiter.

Tensile bond strength of an adhesive resin cement to different alloys having various surface treatments.

PubMed

Abreu, Amara; Loza, Maria A; Elias, Augusto; Mukhopadhyay, Siuli; Looney, Stephen; Rueggeberg, Frederick A

2009-02-01

The ability of a resin cement to bond to a restorative alloy is critical for maximal crown retention to nonideal preparations. Surface treatment and metal type may have an important role in optimizing resin-to-metal strength. The purpose of this study was to examine the effect of surface pretreatment on the tensile strength of base and noble metals bonded using a conventional resin cement. Cylindrical plastic rods (9.5 mm in diameter), cast in base (Rexillium NBF) or noble metal (IPS d.SIGN 53), were divided into rods 10 mm in length (n=10-12). Specimens were heated in a porcelain furnace to create an oxide layer. Test specimens were further subjected to airborne-particle abrasion (50-microm Al(2)O(3) particles) alone or with the application of a metal primer (Alloy Primer). Similarly treated rod ends were joined using resin cement (RelyX ARC), thermocycled (x500, 5 degrees -55 degrees C) and stored (24 hours, 37 degrees C) before debonding using a universal testing machine. Debond strength and failure site were recorded. Rank-based ANOVA for unbalanced designs was used to test for significant interaction (alpha=.050). Each pair of treatments was compared separately for each metal (Bonferroni-adjusted significance level of .0083, overall error rate for comparisons, .05). The 2 metals were compared separately for each of the 3 treatments using an adjusted significance level of .017, maintaining an overall error rate of .05. A multinomial logit model was used to describe the effect of metal type and surface pretreatment on failure site location (alpha=.05). Interaction between metal type and surface pretreatment was significant for stress values (P=.019). Metal type did not significantly affect tensile bond strength for any of the compared surface pretreatments. Metal primer significantly improved tensile bond strength for each metal type. Most failures tended to occur as either adhesive or mixed in nature. Metal primer application significantly enhanced tensile bond strength to base and noble metal. No significant differences in tensile strength were found between alloys. Differences in failure site incidence were found to be related to metal type and surface pretreatment.

Effect of laser power on clad metal in laser-TIG combined metal cladding

NASA Astrophysics Data System (ADS)

Utsumi, Akihiro; Hino, Takanori; Matsuda, Jun; Tasoda, Takashi; Yoneda, Masafumi; Katsumura, Munehide; Yano, Tetsuo; Araki, Takao

2003-03-01

TIG arc welding has been used to date as a method for clad welding of white metal as bearing material. We propose a new clad welding process that combines a CO2 laser and a TIG arc, as a method for cladding at high speed. We hypothesized that this method would permit appropriate control of the melted quantity of base metal by varying the laser power. We carried out cladding while varying the laser power, and investigated the structure near the boundary between the clad layer and the base metal. Using the laser-TIG combined cladding, we found we were able to control appropriately the degree of dilution with the base metal. By applying this result to subsequent cladding, we were able to obtain a clad layer of high quality, which was slightly diluted with the base metal.

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.

Cascaded-cladding-pumped cascaded Raman fiber amplifier.

PubMed

Jiang, Huawei; Zhang, Lei; Feng, Yan

2015-06-01

The conversion efficiency of double-clad Raman fiber laser is limited by the cladding-to-core area ratio. To get high conversion efficiency, the inner-cladding-to-core area ratio has to be less than about 8, which limits the brightness enhancement. To overcome the problem, a cascaded-cladding-pumped cascaded Raman fiber laser with multiple-clad fiber as the Raman gain medium is proposed. A theoretical model of Raman fiber amplifier with multiple-clad fiber is developed, and numerical simulation proves that the proposed scheme can improve the conversion efficiency and brightness enhancement of cladding pumped Raman fiber laser.

Tests for Determining Failure Criteria of Ceramics under Ballistic Impact

DTIC Science & Technology

1992-06-01

strength steel rods, joined together by connectors and pretensioned to remove the slack before prcssurization. The system was designed to operate up...and ftaporta, IJIi J—arfW P*ii Kqfrwar. tv«a 11Q4. Arlington. VA 22201-4)02. an« t« «ha Oflna a« Manag *«** and Rudoat. f«*v«t HatfudtOA rVotoc«(0704-01...different mechanisms for failure and postfailure flow, and these mechanisms operate at different threshold loads. Two one-dimensional computational

Endurance testing of downstream cathodes on a low-power MPD thruster

NASA Technical Reports Server (NTRS)

Burkhart, J. A.; Rose, J. R.

1974-01-01

A low-power MPD thruster with downstream cathode was tested for endurance with a series of hollow cathode designs. Failure modes and failure mechanisms were identified. A new hollow cathode (with rod inserts) has emerged which shows promise for long life. The downstream positioning of the cathode was also changed from an on-axis location to an off-axis location. Data are presented for a 1332-hour life test of this new hollow cathode located at the new off-axis location. Xenon propellant was used.

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

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

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

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

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

DOE PAGES

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

2013-07-12

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

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

Zhang, Chao; Xu, Jun; Cao, Lei

The electrodes of lithium-ion batteries (LIB) are known to be brittle and to fail earlier than the separators during an external crush event. Thus, the understanding of mechanical failure mechanism for LIB electrodes (anode and cathode) is critical for the safety design of LIB cells. In this paper, we present experimental and numerical studies on the constitutive behavior and progression of failure in LIB electrodes. Mechanical tests were designed and conducted to evaluate the constitutive properties of porous electrodes. Constitutive models were developed to describe the stress-strain response of electrodes under uniaxial tensile and compressive loads. The failure criterion andmore » a damage model were introduced to model their unique tensile and compressive failure behavior. The failure mechanism of LIB electrodes was studied using the blunt rod test on dry electrodes, and numerical models were built to simulate progressive failure. The different failure processes were examined and analyzed in detail numerically, and correlated with experimentally observed failure phenomena. Finally, the test results and models improve our understanding of failure behavior in LIB electrodes, and provide constructive insights on future development of physics-based safety design tools for battery structures under mechanical abuse.« less

Effect of Molding and Machining on Neoflon CTFE M400H Polychlorotrifluoroethylene Rod Stock and Valve Seat Properties

NASA Technical Reports Server (NTRS)

Waller, Jess M.; Newton, Barry E.; Beeson, Harold D.

2003-01-01

Since 1997 numerous fires have been reported to the Food and Drug Administration involving cylinder valves installed on medical use oxygen cylinders sold and operated within the United States. All of the cylinder valves in question had polychlorotrifluoroethylene (PCTFE) valve seats. Subsequent failure analysis showed that the main seat was the primary source of ignition. A review of the incidents involving cylinder valve fires indicated three possible ignition mechanisms: contaminant promotion, flow friction, and resonance. However, gas purity analysis showed that uncombusted, residual oxygen was within specification. Infrared and energy dispersive spectroscopy further showed that no contaminants or organic compounds were present in the remaining, uncombusted valve seat material or on seat plug surfaces. Therefore, contaminant-promoted ignition did not appear to be responsible for the failures. Observations of extruded material along the outer edge of the coined or loaded seat area produced by cylinder overuse or poppet overload led to concerns that accelerated gas flow across a deformed seat surface could generate enough localized heating to ignite the polymeric seat. Low molecular weight or highly amorphous quick-quenched PCTFE grades might be expected to be especially prone to this type of deformation. Such a failure mechanism has been described as flow friction; however, the corresponding mechanistic parameters are poorly understood. Subsequent revelation of low-temperature dimensional instability by thermomechanical analysis (TMA) in a variety of PCTFE sheet and rod stock samples led to new concerns that PCTFE valve seats could undergo excessive expansion or contraction during service. During expansion, additional extrusion and accompanying flow friction could occur. During contraction, a gap between the seal and adjacent metal surfaces could form. Gas flowing past the gap could, in turn, lead to resonance heating and subsequent ignition as described in ASTM Guide for Evaluation Nonmetallic Materials for Oxygen Service (G 63). Attempts to uncover the origins of the observed dimensional instability were hindered by uncertainties about resin grade, process history, and post-process heat history introduced by machining, annealing, and sample preparation. An approach was therefore taken to monitor property changes before and after processing and machining using a single, well-characterized lot of Neoflon CTFE.1 M400H resin. A task group consisting of the current PCTFE resin supplier, two molders, and four valve seat manufacturers was formed, and phased testing on raw resin, intermediate rod stock, and finished valve seats initiated. The effect of processing and machining on the properties of PCTFE rod stock and oxygen gas cylinder valve seats was then determined. Testing focused on two types of extruded rod stock and one type of compression-molded rod stock. To accommodate valve seat manufacturer preferences for certain rod stock diameters, two representative diameters were used (4.8 mm (0.1875 in.) and 19.1 mm (0.75 in.)). To encompass a variety of possible sealing configurations, seven different valve seat types with unique geometries or machining histories were tested. The properties investigated were dimensional stability as determined by TMA, specific gravity, differential scanning calorimetry (DSC), compressive strength, zero strength time, and intrinsic viscosity. Findings are discussed in the context of polymer structure-process-property relationships whenever possible.

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

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

The objective of this project is to perform a systematic study of SNF/UNF (spent nuclear fuel/or used nuclear fuel) integrity under simulated transportation environments by using hot cell testing technology developed recently at Oak Ridge National Laboratory (ORNL), CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester). Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmarking tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. With support from the US Department of Energy and the NRC, CIRFT testing has been continued. The CIRFT testing was conducted on three HBR rods (R3, R4,more » and R5), with two specimens failed and one specimen un-failed. The total number of cycles in the test of un-failed specimens went over 2.23 107; the test was stopped as because the specimen did not show any sign of failure. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of used fuel rods in terms of both the curvature amplitude and the maximum of absolute of curvature extremes. The latter is significant because the maxima of extremes signify the maximum of tensile stress of the outer fiber of the bending rod. So far, a large variety of hydrogen contents has been covered in the CIRFT testing on HBR rods. It has been shown that the load amplitude is the dominant factor that controls the lifetime of bending rods, but the hydrogen content also has an important effect on the lifetime attained, according to the load range tested.« less

Experimental Behavior of Fatigued Single Stiffener PRSEUS Specimens

NASA Technical Reports Server (NTRS)

Jegley, Dawn C.

2009-01-01

NASA, the Air Force Research Laboratory and The Boeing Company have worked to develop new low-cost, light-weight composite structures for aircraft. A Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) concept has been developed which offers advantages over traditional metallic structure. In this concept a stitched carbon-epoxy material system has been developed with the potential for reducing the weight and cost of transport aircraft structure by eliminating fasteners, thereby reducing part count and labor. By adding unidirectional carbon rods to the top of stiffeners, the panel becomes more structurally efficient. This combination produces a more damage tolerant design. This document describes the results of experimentation on PRSEUS specimens loaded in unidirectional compression in fatigue and to failure.

Critical cladding radius for hybrid cladding modes

NASA Astrophysics Data System (ADS)

Guyard, Romain; Leduc, Dominique; Lupi, Cyril; Lecieux, Yann

2018-05-01

In this article we explore some properties of the cladding modes guided by a step-index optical fiber. We show that the hybrid modes can be grouped by pairs and that it exists a critical cladding radius for which the modes of a pair share the same electromagnetic structure. We propose a robust method to determine the critical cladding radius and use it to perform a statistical study on the influence of the characteristics of the fiber on the critical cladding radius. Finally we show the importance of the critical cladding radius with respect to the coupling coefficient between the core mode and the cladding modes inside a long period grating.

CONTROL ROD ROTATING MECHANISM

DOEpatents

Baumgarten, A.; Karalis, A.J.

1961-11-28

A threaded rotatable shaft is provided which rotates in response to linear movement of a nut, the shaft being surrounded by a pair of bellows members connected to either side of the nut to effectively seal the reactor from leakage and also to store up energy to shut down the reactor in the event of a power failure. (AEC)

Feasibility of using cone penetrometer truck (CPT) to install time domain reflectometry (TDR) and fiber optic slope failure detectors in pavement structures.

DOT National Transportation Integrated Search

2011-02-01

A new method of cable installation using a heavy-duty Cone Penetration Test : (CPT) truck was developed and practiced successfully in this study. The coaxial and fiber : optic cables were pushed along with the cone rods by the hydraulic system integr...

Lung volumes predict survival in patients with chronic lung allograft dysfunction.

PubMed

Kneidinger, Nikolaus; Milger, Katrin; Janitza, Silke; Ceelen, Felix; Leuschner, Gabriela; Dinkel, Julien; Königshoff, Melanie; Weig, Thomas; Schramm, René; Winter, Hauke; Behr, Jürgen; Neurohr, Claus

2017-04-01

Identification of disease phenotypes might improve the understanding of patients with chronic lung allograft dysfunction (CLAD). The aim of the study was to assess the impact of pulmonary restriction and air trapping by lung volume measurements at the onset of CLAD.A total of 396 bilateral lung transplant recipients were analysed. At onset, CLAD was further categorised based on plethysmography. A restrictive CLAD (R-CLAD) was defined as a loss of total lung capacity from baseline. CLAD with air trapping (AT-CLAD) was defined as an increased ratio of residual volume to total lung capacity. Outcome was survival after CLAD onset. Patients with insufficient clinical information were excluded (n=95).Of 301 lung transplant recipients, 94 (31.2%) developed CLAD. Patients with R-CLAD (n=20) and AT-CLAD (n=21), respectively, had a significantly worse survival (p<0.001) than patients with non-R/AT-CLAD. Both R-CLAD and AT-CLAD were associated with increased mortality when controlling for multiple confounding variables (hazard ratio (HR) 3.57, 95% CI 1.39-9.18; p=0.008; and HR 2.65, 95% CI 1.05-6.68; p=0.039). Furthermore, measurement of lung volumes was useful to identify patients with combined phenotypes.Measurement of lung volumes in the long-term follow-up of lung transplant recipients allows the identification of patients who are at risk for worse outcome and warrant special consideration. Copyright ©ERS 2017.

All fiber cladding mode stripper with uniform heat distribution and high cladding light loss manufactured by CO2 laser ablation

NASA Astrophysics Data System (ADS)

Jebali, M. A.; Basso, E. T.

2018-02-01

Cladding mode strippers are primarily used at the end of a fiber laser cavity to remove high-power excess cladding light without inducing core loss and beam quality degradation. Conventional manufacturing methods of cladding mode strippers include acid etching, abrasive blasting or laser ablation. Manufacturing of cladding mode strippers using laser ablation consist of removing parts of the cladding by fused silica ablation with a controlled penetration and shape. We present and characterize an optimized cladding mode stripper design that increases the cladding light loss with a minimal device length and manufacturing time. This design reduces the localized heat generation by improving the heat distribution along the device. We demonstrate a cladding mode stripper written on a 400um fiber with cladding light loss of 20dB, with less than 0.02dB loss in the core and minimal heating of the fiber and coating. The manufacturing process of the designed component is fully automated and takes less than 3 minutes with a very high throughput yield.

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

Determinants of the biomechanical and radiological outcome of surgical correction of adolescent idiopathic scoliosis surgery: the role of rod properties and patient characteristics.

PubMed

Giudici, Fabrizio; Galbusera, Fabio; Zagra, Antonino; Wilke, Hans-Joachim; Archetti, Marino; Scaramuzzo, Laura

2017-10-01

Aim of the study was to evaluate the role of the mechanical properties of the rod and of the characteristics of the patients (age, skeletal maturity, BMI, and Lenke type) in determining the deformity correction, its maintenance over time and the risk of mechanical failure of the instrumentation. From March 2011 to December 2014 120 patients affected by AIS underwent posterior instrumented fusion. Two 5.5-mm CoCr rods were implanted in all patients. For every patient, age, sex, Risser grade, Lenke type curve, flexibility of the main curve, body mass index (BMI), and percentage of correction were recorded. In all patients, the Cobb angle value and rod curvature angle (RC) were evaluated. RC changes were registered and correlated to each factor to establish a possible statistically significance in a multivariate analysis. A biomechanical model was constructed to study the influence of rod diameter and material as well as the density of the anchoring implants in determining stress and deformation of rods after contouring and implantation. Radiographic and biomechanical analysis showed a different mean rod deformation for concave and convex side: 7.8° and 3.9°, respectively. RC mean value at immediate follow-up was 21.8° for the concave side and 14.6° for the convex. At 2-year minimum follow-up, RC value increases 1.5° only for the concave side. At 3.5-year mean follow-up, RC value increases 2.7°, p = 0.003, for the concave side and 1.3° for the convex, p = 0.06. The use of the stiffest material as well as of the lowest diameter resulted in higher stresses in the rods. The use of either a low or a high instrumentation density resulted only in minor differences in the loss of correction. Rod diameter and material as well as patient characteristics such as BMI, age, and Risser grade play an important role in deformity correction and its maintenance over time.

High temperature gradient cobalt based clad developed using microwave hybrid heating

NASA Astrophysics Data System (ADS)

Prasad, C. Durga; Joladarashi, Sharnappa; Ramesh, M. R.; Sarkar, Anunoy

2018-04-01

The development of cobalt based cladding on a titanium substrate using microwave cladding technique is benchmark in coating area. The developed cladding would serve the function of a corrosion resistant coating under high temperatures. Clads of thickness 500 µm have been developed by microwave hybrid heating. A microwave furnace of 2.45GHz frequency was used at a 900W power level for processing. Impact of processing time on melting and adhesion of clad has been discussed. The study also extended to static thermal analysis of simple parts with cladding using commercial Finite Element analysis (FEA) software. A comparative study is explored between four variants of the clad being developed. The analysis has been conducted using a square sample. Similar temperature gradient is also shown for a proposed multi-layer coating, which includes a thermal barrier coating yttria stabilized zirconia (YSZ) on top of the corrosion resistant clad. The YSZ coating would protect the corrosion resistant cladding and substrate from high temperatures.

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.

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

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

Folsom, Charles P.; Jensen, Colby B.; Williamson, Richard L.

2016-09-01

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure ofmore » the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.« less

High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

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

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

2015-01-01

The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into localmore » stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.« less

Final report of fuel dynamics Test E7

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

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

1977-04-01

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

Biomechanical analysis of occipitocervical stability afforded by three fixation techniques.

PubMed

Helgeson, Melvin D; Lehman, Ronald A; Sasso, Rick C; Dmitriev, Anton E; Mack, Andrew W; Riew, K Daniel

2011-03-01

Occipital condyle screws appear to be a novel technique that demands biomechanical consideration. It has the potential to achieve fixation anterior to the axis of rotation while offering a point of fixation in line with the C1/C2 screws. To compare the segmental stability and range of motion (ROM) of standard occipitocervical (OC) screw/rod and plate constructs versus a new technique that incorporates occipital condyle fixation. Human cadaveric biomechanical analysis. After intact analysis, 10 fresh-frozen human cadaveric OC spine specimens were instrumented bilaterally with C1 lateral mass screws and C2 pedicle screws. Additional occipital instrumentation was tested in random order under the following conditions: standard occipitocervical plate/rod system (Vertex Max; Medtronic, Inc., Minneapolis, MN, USA); occipital condyle screws alone; and occipital condyle screws with the addition of an eyelet screw placed into the occiput bilaterally. After nondestructive ROM testing, specimens were evaluated under computed tomography (CT) and underwent destructive forward flexion failure comparing Group 1 to Group 3. There was no significant difference in OC (Occiput-C1) axial rotation and flexion/extension ROM between the standard occipitocervical plate/rod system (Group 1) and the occipital condyle screws with one eyelet screw bilaterally (Group 3). Furthermore, the occipital condyle screws alone (Group 2) did allow significantly more flexion/extension compared with Group 1. Interestingly, the two groups with occipital condyle screws (Groups 2 and 3) had significantly less lateral bending compared with Group 1. During CT analysis, the mean occipital condyle width was 10.8 mm (range, 9.1-12.7 mm), and the mean condylar length was 24.3 mm (range, 20.2-28.5). On destructive testing, there was no significant difference in forward flexion failure between Groups 1 and 3. With instrumentation across the mobile OC junction, our results indicate that similar stability can be achieved with occipital condyle screws/eyelet screws compared with the standard occipitocervical plate/rod system. Published by Elsevier Inc.

Accident tolerant fuel cladding development: Promise, status, and challenges

NASA Astrophysics Data System (ADS)

Terrani, Kurt A.

2018-04-01

The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

Research on Microstructure and Property of TiC-Co Composite Material Made by Laser Cladding

NASA Astrophysics Data System (ADS)

Zhang, Wei

The experiment of laser cladding on the surface of 2Cr13 steel was made. Titanium carbide (TiC) powder and Co-base alloy powder were used as cladding material. The microstructure and property of laser cladding layer were tested. The research showed that laser cladding layer had better properties such as minute crystals, deeper layer, higher hardness and good metallurgical bonding with base metal. The structure of cladding was supersaturated solid solution with dispersed titanium carbide. The average hardness of cladding zone was 660HV0.2. 2Cr13 steel was widely used in the field of turbine blades. Using laser cladding, the good wear layer would greatly increase the useful life of turbine blades.

Microstructured optical fibers for gas sensing systems

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

Challener, William Albert; Choudhury, Niloy; Palit, Sabarni

2017-10-17

Microstructured optical fiber (MOF) includes a cladding extending a length between first and second ends. The cladding includes an inner porous microstructure that at least partially surrounds a hollow core. A perimeter contour of the hollow core has a non-uniform radial distance from a center axis of the cladding such that first segments of the cladding along the perimeter contour have a shorter radial distance from the center axis relative to second segments of the cladding along the perimeter contour. The cladding receives and propagates light energy through the hollow core, and the inner porous microstructure substantially confines the lightmore » energy within the hollow core. The cladding defines at least one port hole that extends radially from an exterior surface of the cladding to the hollow core. Each port hole penetrates the perimeter contour of the hollow core through one of the second segments of the cladding.« less

SINQ layout, operation, applications and R&D to high power

NASA Astrophysics Data System (ADS)

Bauer, G. S.; Dai, Y.; Wagner, W.

2002-09-01

Since 1997, the Paul Scherrer Institut (PSI) is operating a 1 MW class research spallation neutron source, named SINQ. SINQ is driven by a cascade of three accelerators, the final stage being a 590 MeV isochronous ring cyclotron which delivers a beam current of 1.8 mA at an rf-frequency of 51 MHz. Since for neutron production this is essentially a dc-device, SINQ is a continuous neutron source and is optimized in its design for high time average neutron flux. This makes the facility similar to a research reactor in terms of utilization, but, in terms of beam power, it is, by a large margin, the most powerful spallation neutron source currently in operation world wide. As a consequence, target load levels prevail in SINQ which are beyond the realm of existing experience, demanding a careful approach to the design and operation of a high power target. While the best neutronic performance of the source is expected for a liquid lead-bismuth eutectic target, no experience with such systems exists. For this reason a staged approach has been embarked upon, starting with a heavy water cooled rod target of Zircaloy-2 and proceeding via steel clad lead rods towards the final goal of a target optimised in both, neutronic performance and service life time. Experience currently accruing with a test target containing sample rods with different materials specimens will help to select the proper structural material and make dependable life time estimates accounting for the real operating conditions that prevail in the facility. In parallel, both theoretical and experimental work is going on within the MEGAPIE (MEGAwatt Pilot Experiment) project, a joint initiative by six European research institutions and JAERI (Japan), DOE (USA) and KAERI (Korea), to design, build, operate and explore a liquid lead-bismuth spallation target for 1MW of beam power, taking advantage of the existing spallation neutron facility SINQ.

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

S.O. Bader

The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are releasedmore » from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.« less

Loss of control air at Browns Ferry Unit One: accident sequence analysis

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

Harrington, R.M.; Hodge, S.A.

1986-04-01

This study describes the predicted response of the Browns Ferry Nuclear Plant to a postulated complete failure of plant control air. The failure of plant control air cascades to include the loss of drywell control air at Units 1 and 2. Nevertheless, this is a benign accident unless compounded by simultaneous failures in the turbine-driven high pressure injection systems. Accident sequence calculations are presented for Loss of Control Air sequences with assumed failure upon demand of the Reactor Core Isolation Cooling (RCIC) and the High Pressure Coolant Injection (HPCI) at Unit 1. Sequences with and without operator action are considered.more » Results show that the operators can prevent core uncovery if they take action to utilize the Control Rod Drive Hydraulic System as a backup high pressure injection system.« less

Modelling the behaviour of oxide fuels containing minor actinides with urania, thoria and zirconia matrices in an accelerator-driven system

NASA Astrophysics Data System (ADS)

Sobolev, V.; Lemehov, S.; Messaoudi, N.; Van Uffelen, P.; Aı̈t Abderrahim, H.

2003-06-01

The Belgian Nuclear Research Centre, SCK • CEN, is currently working on the pre-design of the multipurpose accelerator-driven system (ADS) MYRRHA. A demonstration of the possibility of transmutation of minor actinides and long-lived fission products with a realistic design of experimental fuel targets and prognosis of their behaviour under typical ADS conditions is an important task in the MYRRHA project. In the present article, the irradiation behaviour of three different oxide fuel mixtures, containing americium and plutonium - (Am,Pu,U)O 2- x with urania matrix, (Am,Pu,Th)O 2- x with thoria matrix and (Am,Y,Pu,Zr)O 2- x with inert zirconia matrix stabilised by yttria - were simulated with the new fuel performance code MACROS, which is under development and testing at the SCK • CEN. All the fuel rods were considered to be of the same design and sizes: annular fuel pellets, helium bounded with the stainless steel cladding, and a large gas plenum. The liquid lead-bismuth eutectic was used as coolant. Typical irradiation conditions of the hottest fuel assembly of the MYRRHA subcritical core were pre-calculated with the MCNPX code and used in the following calculations as the input data. The results of prediction of the thermo-mechanical behaviour of the designed rods with the considered fuels during three irradiation cycles of 90 EFPD are presented and discussed.

Constitutive behavior and progressive mechanical failure of electrodes in lithium-ion batteries

NASA Astrophysics Data System (ADS)

Zhang, Chao; Xu, Jun; Cao, Lei; Wu, Zenan; Santhanagopalan, Shriram

2017-07-01

The electrodes of lithium-ion batteries (LIB) are known to be brittle and to fail earlier than the separators during an external crush event. Thus, the understanding of mechanical failure mechanism for LIB electrodes (anode and cathode) is critical for the safety design of LIB cells. In this paper, we present experimental and numerical studies on the constitutive behavior and progression of failure in LIB electrodes. Mechanical tests were designed and conducted to evaluate the constitutive properties of porous electrodes. Constitutive models were developed to describe the stress-strain response of electrodes under uniaxial tensile and compressive loads. The failure criterion and a damage model were introduced to model their unique tensile and compressive failure behavior. The failure mechanism of LIB electrodes was studied using the blunt rod test on dry electrodes, and numerical models were built to simulate progressive failure. The different failure processes were examined and analyzed in detail numerically, and correlated with experimentally observed failure phenomena. The test results and models improve our understanding of failure behavior in LIB electrodes, and provide constructive insights on future development of physics-based safety design tools for battery structures under mechanical abuse.

Constitutive behavior and progressive mechanical failure of electrodes in lithium-ion batteries

DOE PAGES

Zhang, Chao; Xu, Jun; Cao, Lei; ...

2017-05-05

The electrodes of lithium-ion batteries (LIB) are known to be brittle and to fail earlier than the separators during an external crush event. Thus, the understanding of mechanical failure mechanism for LIB electrodes (anode and cathode) is critical for the safety design of LIB cells. In this paper, we present experimental and numerical studies on the constitutive behavior and progression of failure in LIB electrodes. Mechanical tests were designed and conducted to evaluate the constitutive properties of porous electrodes. Constitutive models were developed to describe the stress-strain response of electrodes under uniaxial tensile and compressive loads. The failure criterion andmore » a damage model were introduced to model their unique tensile and compressive failure behavior. The failure mechanism of LIB electrodes was studied using the blunt rod test on dry electrodes, and numerical models were built to simulate progressive failure. The different failure processes were examined and analyzed in detail numerically, and correlated with experimentally observed failure phenomena. Finally, the test results and models improve our understanding of failure behavior in LIB electrodes, and provide constructive insights on future development of physics-based safety design tools for battery structures under mechanical abuse.« less

Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

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

Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph

2015-09-01

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T 2O. In a standard processing flowsheet, tritium management would bemore » accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. Samples of Surry-2 and H.B. Robinson pressurized water reactor cladding were heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. Cladding samples were also heated within the temperature range of 480–600ºC expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. The tritium content of the Surry-2 and H.B. Robinson cladding was measured to be ~234 and ~500 µCi/g, respectively. Heating the Surry-2 cladding at 500°C for 24 h removed ~0.2% of the tritium from the cladding, and heating at 700°C for 24 h removed ~9%. Heating the H.B. Robinson cladding at 700°C for 24 h removed ~11% of the tritium. When samples of the Surry-2 and H.B. Robinson claddings were heated at 700°C for 96 h, essentially all of the tritium in the cladding was removed. However, only ~3% of the tritium was removed when a sample of Surry-2 cladding was heated at 600°C for 96 h. These data indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in operations involving cladding recycle.« less

Influence of adding strong-carbide-formation elements multiply on particle-reinforced Fe-matrix composite layer produced by laser cladding

NASA Astrophysics Data System (ADS)

Ma, Mingxing; Liu, Wenjin; Zhong, Minlin; Zhang, Hongjun; Zhang, Weiming

2005-01-01

In the research hotspot of particle reinforced metal-matrix composite layer produced by laser cladding, in-situ reinforced particles obtained by adding strong-carbide-formation elements into cladding power have been attracting more attention for their unique advantage. The research has demonstrated that when adding strong-carbide-formation elements-Ti into the cladding powder of the Fe-C-Si-B separately, by optimizing the composition, better cladding coating with the characters of better strength and toughness, higher wear resistance and free of cracks. When the microstructure of cladding coating is hypoeutectic microstructure, its comprehensive performance is best. The research discovered that, compositely adding the strong-carbide-formation elements like Ti+V, Ti+Zr or V+Zr into the cladding coating is able to improve its comprehensive capability. All the cladding coatings obtained are hypoeutectic microstructure. The cladding coatings have a great deal of particulates, and its average microhardness reaches HV0.2700-1400. The research also discovered that the cladding coating obtained is of less cracking after adding the Ti+Zr.

75 FR 47190 - Airworthiness Directives; Empresa Brasileira de Aeronautica S.A. (EMBRAER) Model EMB-135...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-05

... kind. EMBRAER also states that the rod breakage has not been shown to cause leakage of APU oil in the gearbox, or leakage of the fuel lines in the compartment. EMBRAER states both ignition sources and... detector becomes inoperative. EMBRAER also states that in the event of fire detection failure, [[Page 47191...

Fatigue Analysis of the Piston Rod in a Kaplan Turbine Based on Crack Propagation under Unsteady Hydraulic Loads

NASA Astrophysics Data System (ADS)

Liu, X.; Y Luo, Y.; Wang, Z. W.

2014-03-01

As an important component of the blade-control system in Kaplan turbines, piston rods are subjected to fluctuating forces transferred by the turbines blades from hydraulic pressure oscillations. Damage due to unsteady hydraulic loads might generate unexpected down time and high repair cost. In one running hydropower plant, the fracture failure of the piston rod was found twice at the same location. With the transient dynamic analysis, the retainer ring structure of the piston rod existed a relative high stress concentration. This predicted position of the stress concentration agreed well with the actual fracture position in the plant. However, the local strain approach was not able to explain why this position broke frequently. Since traditional structural fatigue analyses use a local stress strain approach to assess structural integrity, do not consider the effect of flaws which can significantly degrade structural life. Using linear elastic fracture mechanism (LEFM) approaches that include the effect of flaws is becoming common practice in many industries. In this research, a case involving a small semi-ellipse crack was taken into account at the stress concentration area, crack growth progress was calculated by FEM. The relationship between crack length and remaining life was obtained. The crack propagation path approximately agreed with the actual fracture section. The results showed that presence of the crack had significantly changed the local stress and strain distributions of the piston rod compared with non-flaw assumption.

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.

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

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

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

46 CFR 111.60-23 - Metal-clad (Type MC) cable.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b) The...

46 CFR 111.60-23 - Metal-clad (Type MC) cable.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 46 Shipping 4 2011-10-01 2011-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b) The...

Analysis of unclad and sub-clad semi-elliptical flaws in pressure vessel steels

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

Irizarry-Quinones, H.; Macdonald, B.D.; McAfee, W.J.

This study was conducted to support warm prestressing experiments on unclad and sub-clad flawed beams loaded in pure bending. Two cladding yield strengths were investigated: 0.6 Sy and 0.8 Sy, where Sy is the yield strength of the base metal. Cladding and base metal were assumed to be stress free at the stress relief temperature for the 3D elastic-plastic finite element analysis used to model the experiments. The model results indicated that when cooled from the stress relief temperature, the cladding was put in tension due to its greater coefficient of thermal expansion. When cooled, the cladding exhibited various amountsmore » of tensile yielding. The degree of yielding depended on the amount of cooling and the strength of the cladding relative to that of the base metal. When subjected to tensile bending stress, the sub-clad flaw elastic-plastic stress intensity factor, K{sub I}(J), was at first dominated by crack closing force due to tensile yielding in the cladding. Thus, imposed loads initially caused no increase in K{sub I}(J) near the clad-base interface. However, K{sub I}(J) at the flaw depth was little affected. When the cladding residual stress was overcome, K{sub I}(J) gradually increased until the cladding began to flow. Thereafter, the rate at which K{sub I}(J) increased with load was the same as that of an unclad beam. A plastic zone corrected K{sub I} approximation for the unclad flaw was found by the superposition of standard Newman and Raju solutions with those due to a cladding crack closure force approximated by the Kaya and Erdogan solution. These elastic estimates of the effect of cladding in reducing the crack driving force were quite in keeping with the 3D elastic-plastic finite element solution for the sub-clad flaw. The results were also compared with the analysis of clad beam experiments by Keeney and the conclusions by Miyazaki, et al. A number of sub-clad flaw specimens not subjected to warm prestressing were thought to have suffered degraded toughness caused by locally intensified strain aging embrittlement (LISAE) due to welding over the preexisting flaw.« less

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.

OXIDATION OF INCONEL 718 IN AIR AT TEMPERATURES FROM 973K TO 1620K.

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

GREENE,G.A.; FINFROCK,C.C.

2000-10-01

As part of the APT project, it was necessary to quantify the release of tungsten from the APT spallation target during postulated accident conditions in order to develop accident source terms for accident consequence characterization. Experiments with tungsten rods at high temperatures in a flowing steam environment characteristic of postulated accidents revealed that considerable vaporization of the tungsten occurred as a result of reactions with the steam and that the aerosols which formed were readily transported away from the tungsten surfaces, thus exposing fresh tungsten to react with more steam. The resulting tungsten release fractions and source terms were undesirablemore » and it was decided to clad the tungsten target with Inconel 718 in order to protect it from contact with steam during an accident and mitigate the accident source term and the consequences. As part of the material selection criteria, experiments were conducted with Inconel 718 at high temperatures to evaluate the rate of oxidation of the proposed clad material over as wide a temperature range as possible, as well as to determine the high-temperature failure limit of the material. Samples of Inconel 718 were inserted into a preheated furnace at temperatures ranging from 973 K to 1620 K and oxidized in air for varying periods of time. After oxidizing in air at a constant temperature for the prescribed time and then being allowed to cool, the samples would be reweighed to determine their weight gain due to the uptake of oxygen. From these weight gain measurements, it was possible to identify three regimes of oxidation for Inconel 718: a low-temperature regime in which the samples became passivated after the initial oxidation, an intermediate-temperature regime in which the rate of oxidation was limited by diffusion and exhibited a constant parabolic rate dependence, and a high-temperature regime in which material deformation and damage accompanied an accelerated oxidation rate above the parabolic regime. At temperatures below 1173 K, the rate of oxidation of the Inconel 718 surface was found to decrease markedly with time; the parabolic oxidation rate coefficient was not a constant but decreased with time. This was taken to indicate that the oxide film on the surface was having a passivating effect on oxygen transport through the oxide to the underlying metal. For temperatures in the range 1173 K to 1573 K, the time-dependent rate of oxidation as determined once again by weight-gain measurements was found to display the classical parabolic rate behavior, indicating that the rate of transport of reactants through the oxide was controlled by diffusion through the growing oxide layer. Parabolic rate coefficients were determined by least-squares analysis of time-dependent mass-gain data at 1173 K, 1273 K, 1373 K, 1473 K and 1573 K. At temperatures above 1540 K, post test examination of the oxidized samples revealed that the Inconel 718 began to lose strength and to deform. At 1540 K, samples which were suspended from their ends during testing began to demonstrate axial curvature as they lost strength and bowed under their own weight. As the temperatures of the tests were increased, rivulets were seen to appear on the surfaces of the test specimens; damage became severe at 1560 K. Although melting was never observed in any of these tests even up to. 1620 K, it was concluded from these data that the Inconel 718 clad should not be expected to protect the underlying tungsten at temperatures above 1540 K.« less

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

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

NASA Technical Reports Server (NTRS)

Saltsman, J. F.

1973-01-01

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

Influence of laser radiation on structure and properties of steels and alloys

NASA Astrophysics Data System (ADS)

Tarasova, T.; Popova, E.

2013-03-01

In present study, and laser alloying of different steels and laser cladding of Ti and SiC powders mixtures was carried out, and microstructure, as well as microhardness profile and wear properties were examined. Research of the influence of lasers alloying modes on the elastic and plastic characteristics of the surface was conducted. As a result of chemical reactions in the cladded layer, a new phase (TiC) was synthesized during cladding process. The results showed that, in the clad layer, TiC was solidified to form dendrites in the clad zone. Produced coatings have high microhardness values in the upper and middle clad areas, about two time higher than clad matrix microhardness.

Measurement and removal of cladding light in high power fiber systems

NASA Astrophysics Data System (ADS)

Walbaum, Till; Liem, Andreas; Schreiber, Thomas; Eberhardt, Ramona; Tünnermann, Andreas

2018-02-01

The amount of cladding light is important to ensure longevity of high power fiber components. However, it is usually measured either by adding a cladding light stripper (and thus permanently modifying the fiber) or by using a pinhole to only transmit the core light (ignoring that there may be cladding mode content in the core area). We present a novel noninvasive method to measure the cladding light content in double-clad fibers based on extrapolation from a cladding region of constant average intensity. The method can be extended to general multi-layer radially symmetric fibers, e.g. to evaluate light content in refractive index pedestal structures. To effectively remove cladding light in high power systems, cladding light strippers are used. We show that the stripping efficiency can be significantly improved by bending the fiber in such a device and present respective experimental data. Measurements were performed with respect to the numerical aperture as well, showing the dependency of the CLS efficiency on the NA of the cladding light and implying that efficiency data cannot reliably be given for a certain fiber in general without regard to the properties of the guided light.

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

Orientation-Dependent Displacement Sensor Using an Inner Cladding Fiber Bragg Grating.

PubMed

Yang, Tingting; Qiao, Xueguang; Rong, Qiangzhou; Bao, Weijia

2016-09-11

An orientation-dependent displacement sensor based on grating inscription over a fiber core and inner cladding has been demonstrated. The device comprises a short piece of multi-cladding fiber sandwiched between two standard single-mode fibers (SMFs). The grating structure is fabricated by a femtosecond laser side-illumination technique. Two well-defined resonances are achieved by the downstream both core and cladding fiber Bragg gratings (FBGs). The cladding resonance presents fiber bending dependence, together with a strong orientation dependence because of asymmetrical distribution of the "cladding" FBG along the fiber cross-section.

The effect of laser process parameters on microstructure and dilution rate of cladding coatings

NASA Astrophysics Data System (ADS)

Bin, Liu; Heping, Liu; Xingbin, Jing; Yuxin, Li; Peikang, Bai

2018-02-01

In order to broaden the range of application of Q235 steel, it is necessary to repair the surface of steel. High performance 316L stainless steel coating was successfully obtained on Q235 steel by laser cladding technology. The effect of laser cladding parameters on the geometrical size and appearance of single cladding layer was investigated. The experimental results show that laser current has an important influence on the surface morphology of single channel cladding. When the current is from 155A to 165A, the cladding coating becomes smooth. The laser current has an effect on the geometric cross section size and dilution rate of single cladding. The results revealed that with the rising of laser current, the width, height and depth of layer increase gradually. With the rising of laser current, the dilution rate of cladding layer is gradually increasing.

Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2015-12-01

As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

SP-100 multimegawatt scaleup to meet electric propulsion mission requirements

NASA Astrophysics Data System (ADS)

Newkirk, D. W.; Salamah, S. A.; Stewart, S. L.; Pluta, P. R.

The SP-100 nuclear heat source technology, utilizing uranium nitride fuel clad in PWC-11 in a fast reactor with lithium coolant circulated by an electromagnetic pump, is shown to be directly extrapolatable to thermal power levels that meet NASA nuclear electric propulsion requirements using different power conversion techniques. The SP-100 nuclear technology can be applied to missions with NEP (nuclear electric propulsion) requirements as low as tens of kWe to tens of MWe. It is pointed out that the SP-100 heat source has a great advantage of very long lifetime capability, since it utilizes very rugged refractory metal fuel pins and is independent of the power conversion scheme chosen for a given mission. The only moving parts in the nuclear subsystems are the control rods moved to compensate for fuel enrichment degradation due to fission and for power shutdown. Lowest alpha values in the range of interest for potential NASA missions are predicted for the dynamic Rankine and static HYTEC conversion systems.

Recent Progress In Infrared Chalcogenide Glass Fibers

NASA Astrophysics Data System (ADS)

Bornstein, A.; Croitoru, N.; Marom, E.

1984-10-01

Chalcogenide glasses containing elements like As, Ge, Sb and Se have been prepared. A new technique of preparing the raw material and subsequently drawing fibers has been devel-oped in order to avoid the forming of oxygen compounds. The fibers have been drawn by cru-cible and rod method from oxygen free raw material inside an Ar atmosphere glove box. The fibers drawn to date with air and glass cladding have a diameter of 50-500 pm and length of several meterd. Preliminary attenuation measurements indicate that the attentuation is better than 0.1 dB/cm and it is not affected even when the fiber is bent to 2 cm circular radius. The fibes were testes a CO laser beam and were not damaged at power densities below 10 kW/2cm2 CW &100 kw/cm using short pulses 75 n sec. The transmitted power density was 0.8 kW/cm2 which is an appropriate value to the needed for cutting and ablation of human tissues.

Effect of pellet-cladding interaction (PCI) and degradation mechanisms on spent nuclear fuel rod mechanical performance during transportation

NASA Astrophysics Data System (ADS)

Peterson, Brittany Ann

Winter storms can affect millions of people, with impacts such as disruptions to transportation, hazards to human health, reduction in retail sales, and structural damage. Blizzard forecasts for Alberta Clippers can be a particular challenge in the Northern Plains, as these systems typically depart from the Canadian Rockies, intensify, and impact the Northern Plains all within 24 hours. The purpose of this study is to determine whether probabilistic forecasts derived from a local physics-based ensemble can improve specific aspects of winter storm forecasts for three Alberta Clipper cases. Verification is performed on the ensemble members and ensemble mean with a focus on quantifying uncertainty in the storm track, two-meter winds, and precipitation using the MERRA and NOHRSC SNODAS datasets. This study finds that addition improvements are needed to proceed with operational use of the ensemble blizzard products, but the use of a proxy for blizzard conditions yields promising results.

Method and etchant to join ag-clad BSSCO superconducting tape

DOEpatents

Balachandran, Uthamalingam; Iyer, Anand N.; Huang, Jiann Yuan

1999-01-01

A method of removing a silver cladding from high temperature superconducting material clad in silver (HTS) is disclosed. The silver clad HTS is contacted with an aqueous solution of HNO.sub.3 followed by an aqueous solution of NH.sub.4 OH and H.sub.2 O.sub.2 for a time sufficient to remove the silver cladding from the superconducting material without adversely affecting the superconducting properties of the superconducting material. A portion of the silver cladding may be masked with a material chemically impervious to HNO.sub.3 and to a combination of NH.sub.4 OH and H.sub.2 O.sub.2 to preserve the Ag coating. A silver clad superconductor is disclosed, made in accordance with the method discussed.

Formation of anomalous eutectic in Ni-Sn alloy by laser cladding

NASA Astrophysics Data System (ADS)

Wang, Zhitai; Lin, Xin; Cao, Yongqing; Liu, Fencheng; Huang, Weidong

2018-02-01

Ni-Sn anomalous eutectic is obtained by single track laser cladding with the scanning velocity from 1 mm/s to 10 mm/s using the Ni-32.5 wt.%Sn eutectic powders. The microstructure of the cladding layer and the grain orientations of anomalous eutectic were investigated. It is found that the microstructure is transformed from primary α-Ni dendrites and the interdendritic (α-Ni + Ni3Sn) eutectic at the bottom of the cladding layer to α-Ni and β-Ni3Sn anomalous eutectic at the top of the cladding layer, whether for single layer or multilayer laser cladding. The EBSD maps and pole figures indicate that the spatially structure of α-Ni phase is discontinuous and the Ni3Sn phase is continuous in anomalous eutectic. The transformation from epitaxial growth columnar at bottom of cladding layer to free nucleation equiaxed at the top occurs, i.e., the columnar to equiaxed transition (CET) at the top of cladding layer during laser cladding processing leads to the generation of anomalous eutectic.

Laser welding of a cobalt-chromium removable partial denture alloy.

PubMed

NaBadalung, D P; Nicholls, J I

1998-03-01

The electric alloy brazed joints of removable partial denture alloys have failed frequently after routine usage. A technique providing higher joint strengths was investigated. This investigation compared the tensile strengths of electric-brazed and laser-welded joints for a cobalt-chromium removable partial denture alloy. Twenty-four cobalt-chromium standard tensile testing rods were prepared and divided into three groups of eight. All specimens in the control group (group 1) were left in the as-cast condition. Groups 2 and 3 were the test specimens, which were sectioned at the center of the rod. Eight specimens were joined by using electric brazing, and the remaining specimens were joined by using laser welding. After joining, each joint was ground to a uniform diameter, then tested to tensile failure on an Instron universal testing machine. Failure loads were recorded and fracture stress calculated. Statistical analysis was applied. The student-Newman-Keuls test showed a highly significant difference between the joint strengths of the as-cast control specimens, the electric-brazed and laser-welded joints. The tensile strengths of the as-cast joints were higher than those for the laser-welded joints, and both were higher than the electric-brazed joint strengths.

Influence of the Manufacturing Process on Defects in the Galvanized Coating of High Carbon Steel Wires.

PubMed

Gelfi, Marcello; Solazzi, Luigi; Poli, Sandro

2017-03-06

This study is a detailed failure analysis of galvanized high carbon steel wires, which developed coating cracks during the torsion test performed as a quality control at the end of the manufacturing process. Careful visual inspections showed that the cracks are already present in the coating before the torsion test. In order to explain the origin of these cracks, systematic metallographic investigations were performed by means of optical and scanning electron microscope on both the wires and the rods that have been cold drawn to produce the wire. The chemical composition of the galvanized coatings was evaluated by means of energy dispersive spectroscopy. Micro bidimensional X-ray diffraction experiments were also performed to measure the residual stresses in the galvanized coating. The results showed that the failure is related to two main factors: the relatively high content of silicon in the steel and the unsuitable cooling rate of the rods at the exit from the galvanizing bath. The mechanism proposed to explain the origin of the defects was supported by Finite Elements Methods simulations and verified with in-plant tests. The proper countermeasures were then applied and the problem successfully solved.

An interesting case of rifampicin-dependent/-enhanced multidrug-resistant tuberculosis.

PubMed

Zhong, M; Zhang, X; Wang, Y; Zhang, C; Chen, G; Hu, P; Li, M; Zhu, B; Zhang, W; Zhang, Y

2010-01-01

We report a case of rifampicin (RMP) dependent/enhanced multidrug-resistant (MDR-TB) from a patient who had been treated with the World Health Organization optional thrice-weekly treatment and document the clinical and bacteriological features. RMP-enhanced tubercle bacilli that grew poorly without RMP but grew better in its presence were isolated from the patient with treatment failure. The bacteria grown without RMP consisted of mixed morphologies of short rod-shaped acid-fast bacteria and poorly stained coccoid bacteria, but stained normally as acid-fast rods when grown in the presence of RMP. The isolated RMP-enhanced bacteria harbored the common S531L mutation and a novel mutation F584S in the rpoB gene. Treatment containing RMP or replacement of RMP with more powerful rifapentine paradoxically aggravated the disease, but its removal led to successful cure of the patient. This study highlights the potential dangers of continued treatment of MDR-TB with rifamycins that can occur due to delayed or absent drug susceptibility results and calls for timely detection of RMP-dependent/-enhanced bacteria in treatment failure patients by including RMP in culture media and removal of RMP from treatment regimen upon detection.

Influence of the Manufacturing Process on Defects in the Galvanized Coating of High Carbon Steel Wires

PubMed Central

Gelfi, Marcello; Solazzi, Luigi; Poli, Sandro

2017-01-01

This study is a detailed failure analysis of galvanized high carbon steel wires, which developed coating cracks during the torsion test performed as a quality control at the end of the manufacturing process. Careful visual inspections showed that the cracks are already present in the coating before the torsion test. In order to explain the origin of these cracks, systematic metallographic investigations were performed by means of optical and scanning electron microscope on both the wires and the rods that have been cold drawn to produce the wire. The chemical composition of the galvanized coatings was evaluated by means of energy dispersive spectroscopy. Micro bidimensional X-ray diffraction experiments were also performed to measure the residual stresses in the galvanized coating. The results showed that the failure is related to two main factors: the relatively high content of silicon in the steel and the unsuitable cooling rate of the rods at the exit from the galvanizing bath. The mechanism proposed to explain the origin of the defects was supported by Finite Elements Methods simulations and verified with in-plant tests. The proper countermeasures were then applied and the problem successfully solved. PMID:28772623

Multi-clad black display panel

DOEpatents

Veligdan, James T.; Biscardi, Cyrus; Brewster, Calvin

2002-01-01

A multi-clad black display panel, and a method of making a multi-clad black display panel, are disclosed, wherein a plurality of waveguides, each of which includes a light-transmissive core placed between an opposing pair of transparent cladding layers and a black layer disposed between transparent cladding layers, are stacked together and sawed at an angle to produce a wedge-shaped optical panel having an inlet face and an outlet face.

Nuclear-powered pacemaker fuel cladding study. [Difficulty of dissolving cladding and /sup 238/PuO/sub 2/ for obtaining materials for acts of terrorism

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

Shoup, R.L.

1976-07-01

The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a /sup 238/PuO/sub 2/-powered pacemaker could be transformed into a terrorism weapon.

Research on microstructure properties of the TiC/Ni-Fe-Al coating prepared by laser cladding technology

NASA Astrophysics Data System (ADS)

Jiao, Junke; Xu, Zifa; Zan, Shaoping; Zhang, Wenwu; Sheng, Liyuan

2017-10-01

In this paper, the laser cladding method was used to preparation the TiC reinforced Ni-Fe-Al coating on the Ni base superalloy. The Ti/Ni-Fe-Al powder was preset on the Ni base superalloy and the powder layer thickness is 0.5mm. A fiber laser was used the melting Ti/Ni-Fe-Al powder in an inert gas environment. The shape of the cladding layer was tested using laser scanning confocal microscope (LSCM) under different cladding parameters such as the laser power, the melting velocity and the defocused amount. The microstructure, the micro-hardness was tested by LSCM, SEM, Vickers hardness tester. The test result showed that the TiC particles was distributed uniformly in the cladding layer and hardness of the cladding layer was improved from 180HV to 320HV compared with the Ni-Fe-Al cladding layer without TiC powder reinforced, and a metallurgical bonding was produced between the cladding layer and the base metal. The TiC powder could make the Ni-Fe-Al cladding layer grain refining, and the more TiC powder added in the Ni-Fe-Al powder, the smaller grain size was in the cladding layer.

Laser doppler and radar interferometer for contactless measurements on unaccessible tie-rods on monumental buildings: Santa Maria della Consolazione Temple in Todi

NASA Astrophysics Data System (ADS)

Gioffré, M.; Cavalagli, N.; Pepi, C.; Trequattrini, M.

2017-01-01

Non-contact measurements can be effectively used in civil engineering to assess the variation of structural performance with time. In the last decades this approach has received considerable interests from researchers working in the field of structural health monitoring (SHM). Indeed, non-contact measurements are very attractive because it is possible to perform non intrusive and non destructive investigations even being at a significant distance from the targets. Within this context, contactless measurements of the tie-rod vibrations in the Santa Maria della Consolazione Temple in Todi (Italy) are presented in this paper. In particular, laser vibrometer and radar interferometer measurements are used to estimate natural frequencies and mode shapes. This information is crucial to obtain the tensile axial force in the tie-rods, which can be used as an indicator of structural integrity or possible failure. Furthermore, a novel approach is proposed where drones (Unmanned Aerial Vehicles) can be successfully used to improve the effectiveness and the accuracy of the experimental activities.

Method and etchant to join Ag-clad BSSCO superconducting tape

DOEpatents

Balachandran, U.; Iyer, A.N.; Huang, J.Y.

1999-03-16

A method of removing a silver cladding from high temperature superconducting material clad in silver (HTS) is disclosed. The silver clad HTS is contacted with an aqueous solution of HNO{sub 3} followed by an aqueous solution of NH{sub 4}OH and H{sub 2}O{sub 2} for a time sufficient to remove the silver cladding from the superconducting material without adversely affecting the superconducting properties of the superconducting material. A portion of the silver cladding may be masked with a material chemically impervious to HNO{sub 3} and to a combination of NH{sub 4}OH and H{sub 2}O{sub 2} to preserve the Ag coating. A silver clad superconductor is disclosed, made in accordance with the method discussed. 3 figs.

In-situ material-motion diagnostics and fuel radiography in experimental reactors

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

DeVolpi, A.

1982-01-01

Material-motion monitoring has become a routine part of in-pile transient reactor-safety experiments. Diagnostic systems, such as the fast-neutron hodoscope, were developed for the purpose of providing direct time-resolved data on pre-failure fuel motion, cladding-breach time and location, and post-failure fuel relocation. Hodoscopes for this purpose have been installed at TREAT and CABRI; other types of imaging systems that have been tested are a coded-aperture at ACRR and a pinhole at TREAT. Diagnostic systems that use penetrating radiation emitted from the test section can non-invasively monitor fuel without damage to the measuring instrument during the radiographic images of test sections installedmore » in the reator. Studies have been made of applications of hodoscopes to other experimental reactors, including PBF, FARET, STF, ETR, EBR-II, SAREF-STF, and DMT.« less

Early implementation of SiC cladding fuel performance models in BISON

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

Powers, Jeffrey J.

2015-09-18

SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation duemore » to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.« less

Strengthening Effect of Incremental Shear Deformation on Ti Alloy Clad Plate with a Ni-Based Alloy Laser-Clad Layer

NASA Astrophysics Data System (ADS)

Zhao, W.; Zha, G. C.; Kong, F. X.; Wu, M. L.; Feng, X.; Gao, S. Y.

2017-05-01

A Ti-6Al-4V alloy clad plate with a Tribaloy 700 alloy laser-clad layer is subjected to incremental shear deformation, and we evaluate the structural evolution and mechanical properties of the specimens. Results indicate the significance of the incremental shear deformation on the strengthening effect. The wear resistance and Vickers hardness of the laser-clad layer are enhanced due to increased dislocation density. The incremental shear deformation can increase the bonding strength of the laser-clad layer and the corresponding substrate and can break the columnar crystals in the laser-clad layer near the interface. These phenomena suggest that shear deformation eliminates the defects on the interface of the laser-clad layer and the substrate. Substrate hardness is evidently improved, and the strengthening effect is caused by the increased dislocation density and shear deformation. This deformation can then transform the α- and β-phases in the substrate into a high-intensity ω-phase.

Orientation-dependent fiber-optic accelerometer based on grating inscription over fiber cladding.

PubMed

Rong, Qiangzhou; Qiao, Xueguang; Guo, Tuan; Bao, Weijia; Su, Dan; Yang, Hangzhou

2014-12-01

An orientation-sensitive fiber-optic accelerometer based on grating inscription over fiber cladding has been demonstrated. The sensor probe comprises a compact structure in which a short section of thin-core fiber (TCF) stub containing a "cladding" fiber Bragg grating (FBG) is spliced to another single-mode fiber (SMF) without any lateral offset. A femtosecond laser side-illumination technique was utilized to ensure that the grating inscription remains close to the core-cladding interface of the TCF. The core mode and the cladding mode of the TCF are coupled at the core-mismatch junction, and two well-defined resonances in reflection appear from the downstream FBG, in which the cladding resonance exhibits a strong polarization and bending dependence due to the asymmetrical distribution of the cladding FBG along the fiber cross section. Strong orientation dependence of the vibration (acceleration) measurement has been achieved by power detection of the cladding resonance. Meanwhile, the unwanted power fluctuations and temperature perturbations can be referenced out by monitoring the fundamental core resonance.

Evolution of transmission spectra of double cladding fiber during etching

NASA Astrophysics Data System (ADS)

Ivanov, Oleg V.; Tian, Fei; Du, Henry

2017-11-01

We investigate the evolution of optical transmission through a double cladding fiber-optic structure during etching. The structure is formed by a section of SM630 fiber with inner depressed cladding between standard SMF-28 fibers. Its transmission spectrum exhibits two resonance dips at wavelengths where two cladding modes have almost equal propagation constants. We measure transmission spectra with decreasing thickness of the cladding and show that the resonance dips shift to shorter wavelengths, while new dips of lower order modes appear from long wavelength side. We calculate propagation constants of cladding modes and resonance wavelengths, which we compare with the experiment.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

NASA Astrophysics Data System (ADS)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

2015-06-01

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

Capturing reflected cladding modes from a fiber Bragg grating with a double-clad fiber coupler.

PubMed

Baiad, Mohamad Diaa; Gagné, Mathieu; Lemire-Renaud, Simon; De Montigny, Etienne; Madore, Wendy-Julie; Godbout, Nicolas; Boudoux, Caroline; Kashyap, Raman

2013-03-25

We present a novel measurement scheme using a double-clad fiber coupler (DCFC) and a fiber Bragg grating (FBG) to resolve cladding modes. Direct measurement of the optical spectra and power in the cladding modes is obtained through the use of a specially designed DCFC spliced to a highly reflective FBG written into slightly etched standard photosensitive single mode fiber to match the inner cladding diameter of the DCFC. The DCFC is made by tapering and fusing two double-clad fibers (DCF) together. The device is capable of capturing backward propagating low and high order cladding modes simply and efficiently. Also, we demonstrate the capability of such a device to measure the surrounding refractive index (SRI) with an extremely high sensitivity of 69.769 ± 0.035 μW/RIU and a resolution of 1.433 × 10(-5) ± 8 × 10(-9) RIU between 1.37 and 1.45 RIU. The device provides a large SRI operating range from 1.30 to 1.45 RIU with sufficient discrimination for all individual captured cladding modes. The proposed scheme can be adapted to many different types of bend, temperature, refractive index and other evanescent wave based sensors.

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

DOE PAGES

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; ...

2016-07-15

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

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

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Finite Element Analysis of Laser Engineered Net Shape (LENS™) Tungsten Clad Squeeze Pins

NASA Astrophysics Data System (ADS)

Sakhuja, Amit; Brevick, Jerald R.

2004-06-01

In the aluminum high-pressure die-casting and indirect squeeze casting processes, local "squeeze" pins are often used to minimize internal solidification shrinkage in heavy casting sections. Squeeze pins frequently fail in service due to molten aluminum adhering to the H13 tool steel pins ("soldering"). A wide variety of coating materials and methods have been developed to minimize soldering on H13. However, these coatings are typically very thin, and experience has shown their performance on squeeze pins is highly variable. The LENS™ process was employed in this research to deposit a relatively thick tungsten cladding on squeeze pins. An advantage of this process was that the process parameters could be precisely controlled in order to produce a satisfactory cladding. Two fixtures were designed and constructed to enable the end and outer diameter (OD) of the squeeze pins to be clad. Analyses were performed on the clad pins to evaluate the microstructure and chemical composition of the tungsten cladding and the cladding-H13 substrate interface. A thermo-mechanical finite element analysis (FEA) was performed to assess the stress distribution as a function of cladding thickness on the pins during a typical casting thermal cycle. FEA results were validated via a physical test, where the clad squeeze pins were immersed into molten aluminum. Pins subjected to the test were evaluated for thermally induced cracking and resistance to soldering of the tungsten cladding.

Transmission of laser pulses with high output beam quality using step-index fibers having large cladding

DOEpatents

Yalin, Azer P; Joshi, Sachin

2014-06-03

An apparatus and method for transmission of laser pulses with high output beam quality using large core step-index silica optical fibers having thick cladding, are described. The thick cladding suppresses diffusion of modal power to higher order modes at the core-cladding interface, thereby enabling higher beam quality, M.sup.2, than are observed for large core, thin cladding optical fibers. For a given NA and core size, the thicker the cladding, the better the output beam quality. Mode coupling coefficients, D, has been found to scale approximately as the inverse square of the cladding dimension and the inverse square root of the wavelength. Output from a 2 m long silica optical fiber having a 100 .mu.m core and a 660 .mu.m cladding was found to be close to single mode, with an M.sup.2=1.6. Another thick cladding fiber (400 .mu.m core and 720 .mu.m clad) was used to transmit 1064 nm pulses of nanosecond duration with high beam quality to form gas sparks at the focused output (focused intensity of >100 GW/cm.sup.2), wherein the energy in the core was <6 mJ, and the duration of the laser pulses was about 6 ns. Extending the pulse duration provided the ability to increase the delivered pulse energy (>20 mJ delivered for 50 ns pulses) without damaging the silica fiber.

Robust cladding light stripper for high-power fiber lasers using soft metals.

PubMed

Babazadeh, Amin; Nasirabad, Reza Rezaei; Norouzey, Ahmad; Hejaz, Kamran; Poozesh, Reza; Heidariazar, Amir; Golshan, Ali Hamedani; Roohforouz, Ali; Jafari, S Naser Tabatabaei; Lafouti, Majid

2014-04-20

In this paper we present a novel method to reliably strip the unwanted cladding light in high-power fiber lasers. Soft metals are utilized to fabricate a high-power cladding light stripper (CLS). The capability of indium (In), aluminum (Al), tin (Sn), and gold (Au) in extracting unwanted cladding light is examined. The experiments show that these metals have the right features for stripping the unwanted light out of the cladding. We also find that the metal-cladding contact area is of great importance because it determines the attenuation and the thermal load on the CLS. These metals are examined in different forms to optimize the contact area to have the highest possible attenuation and avoid localized heating. The results show that sheets of indium are very effective in stripping unwanted cladding light.

Impact of forced vital capacity loss on survival after the onset of chronic lung allograft dysfunction.

PubMed

Todd, Jamie L; Jain, Rahil; Pavlisko, Elizabeth N; Finlen Copeland, C Ashley; Reynolds, John M; Snyder, Laurie D; Palmer, Scott M

2014-01-15

Emerging evidence suggests a restrictive phenotype of chronic lung allograft dysfunction (CLAD) exists; however, the optimal approach to its diagnosis and clinical significance is uncertain. To evaluate the hypothesis that spirometric indices more suggestive of a restrictive ventilatory defect, such as loss of FVC, identify patients with distinct clinical, radiographic, and pathologic features, including worse survival. Retrospective, single-center analysis of 566 consecutive first bilateral lung recipients transplanted over a 12-year period. A total of 216 patients developed CLAD during follow-up. CLAD was categorized at its onset into discrete physiologic groups based on spirometric criteria. Imaging and histologic studies were reviewed when available. Survival after CLAD diagnosis was assessed using Kaplan-Meier and Cox proportional hazards models. Among patients with CLAD, 30% demonstrated an FVC decrement at its onset. These patients were more likely to be female, have radiographic alveolar or interstitial changes, and histologic findings of interstitial fibrosis. Patients with FVC decline at CLAD onset had significantly worse survival after CLAD when compared with those with preserved FVC (P < 0.0001; 3-yr survival estimates 9% vs. 48%, respectively). The deleterious impact of CLAD accompanied by FVC loss on post-CLAD survival persisted in a multivariable model including baseline demographic and clinical factors (P < 0.0001; adjusted hazard ratio, 2.73; 95% confidence interval, 1.86-4.04). At CLAD onset, a subset of patients demonstrating physiology more suggestive of restriction experience worse clinical outcomes. Further study of the biologic mechanisms underlying CLAD phenotypes is critical to improving long-term survival after lung transplantation.

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

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Kazimi, Mujid S.

2013-10-01

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

Long-term outcomes and management of lung transplant recipients.

PubMed

Costa, Joseph; Benvenuto, Luke J; Sonett, Joshua R

2017-06-01

Lung transplantation is an established treatment for patients with end-stage lung disease. Improvements in immunosuppression and therapeutic management of infections have resulted in improved long-term survival and a decline in allograft rejection. Allograft rejection continues to be a serious complication following lung transplantation, thereby leading to acute graft failure and, subsequently, chronic lung allograft dysfunction (CLAD). Bronchiolitis obliterans syndrome (BOS), the most common phenotype of CLAD, is the leading cause of late mortality and morbidity in lung recipients, with 50% having developed BOS within 5 years of lung transplantation. Infections in lung transplant recipients are also a significant complication and represent the most common cause of death within the first year. The success of lung transplantation depends on careful management of immunosuppressive regimens to reduce the rate of rejection, while monitoring recipients for infections and complications to help identify problems early. The long-term outcomes and management of lung transplant recipients are critically based on modulating natural immune response of the recipient to prevent acute and chronic rejection. Understanding the immune mechanisms and temporal correlation of acute and chronic rejection is thus critical in the long-term management of lung recipients. Copyright © 2017 Elsevier Ltd. All rights reserved.

Improved gas tagging and cover gas combination for nuclear reactor

DOEpatents

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

1983-09-26

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

Critical Safe Disposal of Spent Fuel: Behavior of Neutron Poisons

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

Kienzler, Bernhard; Gmal, Bernhard

2007-07-01

In contrast to Yucca Mountain, European repository concepts rely on deep underground conditions which guarantee permanently a reducing geochemical environment. As long as no water comes into contact with the disposed nuclear fuel, criticality is excluded by compliance with the disposal conditions (limitation of U/Pu in the canisters). Penetration of water into the canister may also be considered as a scenario. However, water in a disposal results in geochemical reactions proceeding over very long periods of time: (1) Presence of water allows the corrosion of the steel of the canister material forming hydrogen and iron corrosion products. (2) Hydrogen pressuresmore » affect the zircaloy cladding even at low temperatures. Failure of fuel cladding and spacers leads to changes in the geometrical configuration. (3) UO{sub 2} matrix corrosion results in geochemically controlled reformation of secondary phase. (4) Even if the dissolution rate of UO{sub 2} is low, elements accounting for burnup credit do not behave similar as uranium. Geochemical reactions are analyzed in detail and compositions are presented which have a high probability to be formed in the long-term needing to be analyzed with respect to K{sub eff}. (authors)« less

Automated power control system for reactor TRIGA PUSPATI

NASA Astrophysics Data System (ADS)

Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

2017-01-01

Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

Electromagnetic Calculation of Combined Earthing System with Ring Earth Electrode and Vertical Rods for Wind Turbine

NASA Astrophysics Data System (ADS)

Fujii, Toshiaki; Yasuda, Yoh; Ueda, Toshiaki

With the worldwide spread of wind turbine installations, various problems such as landscape issues, bird strikes and grid connections have arisen. Protection of wind turbines from lightning is cited as one of the main problems. Wind turbines are often struck by lightning because of their open-air locations, such as in mountainous areas, and their special configuration and very-high construction. Especially, low-voltage and control circuits can fail or suffer burnout while blades can incur serious damage if struck by lightning. Wind turbine failures caused by lightning strikes account for approximately 25% of all failures. The problem is regarded as a global one that needs immediate resolution. It is important to understand the impedance characteristics of wind turbine earthing systems from the viewpoint of lightning protection. A report from IEC TR61400-24 recommends a “ring earth electrode”. This was originally defined in IEC 61024 (currently revised and re-numbered as IEC 62305), where such an electrode is recommended to reduce touch and step voltages in households and buildings. IEC TR61400-24 also recommended additional electrodes of vertical or horizontal rods. However, these concepts have not been fully discussed from the viewpoint of its application to wind turbines. To confirm the effect of a combination of a ring earth electrode and additional vertical rods for protection of a wind turbine, this report uses the Finite Difference Time Domain (FDTD) method to present an electromagnetic transient analysis on such a wind turbine earthing system. The results show that an optimal combination can be arranged from viewpoints of lightning protection and construction cost. Thus, this report discusses how to establish a quantitative design methodology of the wind turbine earthing system to provide effective lightning protection.

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

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

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

NASA Astrophysics Data System (ADS)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2017-12-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

NASA Astrophysics Data System (ADS)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2018-02-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Examination of UC-ZrC after long term irradiation at thermionic temperature

NASA Technical Reports Server (NTRS)

Yang, L.; Johnson, H. O.

1972-01-01

Two fluoride tungsten clad UC-ZrC fueled capsules, designated as V-2C and V-2D, were examined a hot cell after irradiation in NASA Plum Brook Reactor at a maximum cladding temperature of 1930 K for 11,089 and 12,031 hours to burnups of 3.0 x 10 to the 20th power and 2.1 x 10 to the 20th power fission/c.c. respectively. Percentage of fission gas release from the fuel material was measured by radiochemical means. Cladding deformation, fuel-cladding interaction and microstructures of fuel, cladding, and fuel-cladding interface were studied metallographically. Compositions of dispersions in fuel, fuel matrix and fuel-cladding interaction layer were analyzed by electron microprobe techniques. Axial and radial distributions of burnup were determined by gamma-scan, autoradiography and isotopic burnup analysis. The results are presented and discussed in conjunction with the requirements of thermionic fuel elements for space power application.

CXCL4 Contributes to the Pathogenesis of Chronic Liver Allograft Dysfunction

PubMed Central

Li, Jing; Shi, Yuan; Xie, Ke-Liang; Yin, Hai-Fang; Yan, Lu-nan; Lau, Wan-yee; Wang, Guo-Lin

2016-01-01

Chronic liver allograft dysfunction (CLAD) remains the most common cause of patient morbidity and allograft loss in liver transplant patients. However, the pathogenesis of CLAD has not been completely elucidated. By establishing rat CLAD models, in this study, we identified the informative CLAD-associated genes using isobaric tags for relative and absolute quantification (iTRAQ) proteomics analysis and validated these results in recipient rat liver allografts. CXCL4, CXCR3, EGFR, JAK2, STAT3, and Collagen IV were associated with CLAD pathogenesis. We validated that CXCL4 is upstream of these informative genes in the isolated hepatic stellate cells (HSC). Blocking CXCL4 protects against CLAD by reducing liver fibrosis. Therefore, our results indicated that therapeutic approaches that neutralize CXCL4, a newly identified target of fibrosis, may represent a novel strategy for preventing and treating CLAD after liver transplantation. PMID:28053995

CXCL4 Contributes to the Pathogenesis of Chronic Liver Allograft Dysfunction.

PubMed

Li, Jing; Liu, Bin; Shi, Yuan; Xie, Ke-Liang; Yin, Hai-Fang; Yan, Lu-Nan; Lau, Wan-Yee; Wang, Guo-Lin

2016-01-01

Chronic liver allograft dysfunction (CLAD) remains the most common cause of patient morbidity and allograft loss in liver transplant patients. However, the pathogenesis of CLAD has not been completely elucidated. By establishing rat CLAD models, in this study, we identified the informative CLAD-associated genes using isobaric tags for relative and absolute quantification (iTRAQ) proteomics analysis and validated these results in recipient rat liver allografts. CXCL4, CXCR3, EGFR, JAK2, STAT3, and Collagen IV were associated with CLAD pathogenesis. We validated that CXCL4 is upstream of these informative genes in the isolated hepatic stellate cells (HSC). Blocking CXCL4 protects against CLAD by reducing liver fibrosis. Therefore, our results indicated that therapeutic approaches that neutralize CXCL4, a newly identified target of fibrosis, may represent a novel strategy for preventing and treating CLAD after liver transplantation.

Multimode optical fiber

DOEpatents

Bigot-Astruc, Marianne; Molin, Denis; Sillard, Pierre

2014-11-04

A depressed graded-index multimode optical fiber includes a central core, an inner depressed cladding, a depressed trench, an outer depressed cladding, and an outer cladding. The central core has an alpha-index profile. The depressed claddings limit the impact of leaky modes on optical-fiber performance characteristics (e.g., bandwidth, core size, and/or numerical aperture).

Femtosecond-laser inscribed double-cladding waveguides in Nd:YAG crystal: a promising prototype for integrated lasers.

PubMed

Liu, Hongliang; Chen, Feng; Vázquez de Aldana, Javier R; Jaque, D

2013-09-01

We report on the design and implementation of a prototype of optical waveguides fabricated in Nd:YAG crystals by using femtosecond-laser irradiation. In this prototype, two concentric tubular structures with nearly circular cross sections of different diameters have been inscribed in the Nd:YAG crystals, generating double-cladding waveguides. Under 808 nm optical pumping, waveguide lasers have been realized in the double-cladding structures. Compared with single-cladding waveguides, the concentric tubular structures, benefiting from the large pump area of the outermost cladding, possess both superior laser performance and nearly single-mode beam profile in the inner cladding. Double-cladding waveguides of the same size were fabricated and coated by a thin optical film, and a maximum output power of 384 mW and a slope efficiency of 46.1% were obtained. Since the large diameters of the outer claddings are comparable with those of the optical fibers, this prototype paves a way to construct an integrated single-mode laser system with a direct fiber-waveguide configuration.

Microstructure and properties of Fe-based composite coating by laser cladding Fe-Ti-V-Cr-C-CeO2 powder

NASA Astrophysics Data System (ADS)

Zhang, Hui; Zou, Yong; Zou, Zengda; Wu, Dongting

2015-01-01

In situ TiC-VC reinforced Fe-based cladding layer was obtained on low carbon steel surface by laser cladding with Fe-Ti-V-Cr-C-CeO2 alloy powder. The microstructure, phases and properties of the cladding layer were investigated by X-ray diffractometry (XRD), scanning electron microscopy (SEM), energy dispersive spectrometry (EDS), transmission electron microscopy (TEM), potentio-dynamic polarization and electro-chemical impedance spectroscopy (EIS). Results showed Fe-Ti-V-Cr-C-CeO2 alloy powder formed a good cladding layer without defects such as cracks and pores. The phases of the cladding layer were α-Fe, γ-Fe, TiC, VC and TiVC2. The microstructures of the cladding layer matrix were lath martensite and retained austenite. The carbides were polygonal blocks with a size of 0.5-2 μm and distributed uniformly in the cladding layer. High resolution transmission electron microscopy showed the carbide was a complex matter composed of nano TiC, VC and TiVC2. The cladding layer with a hardness of 1030 HV0.2 possessed good wear and corrosion resistance, which was about 16.85 and 9.06 times than that of the substrate respectively.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

DOE PAGES

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; ...

2015-03-19

FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory insidemore » the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.« less

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

Characterizing the effects of cladding on semi-elliptical longitudinal surface flaws in cylindrical vessels subjected to internal pressure

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

Killian, D.E.; Yoon, K.K.

1996-12-01

Flaws on the inside surface of cladded reactor vessels are often analyzed by modelling the carbon steel base metal without consideration of a layer of stainless steel cladding material, thus ignoring the effects of this bimetallic discontinuity. Adding cladding material to the inside surface of a finite element model of a vessel raises concerns regarding adequate mesh refinement in the vicinity of the base metal/cladding interface. This paper presents results of three-dimensional linear stress analysis that has been performed to obtain stress intensity factors for clad and unclad reactor vessels subjected to internal pressure loading. The study concentrates on semi-ellipticalmore » longitudinal surface flaws with a 6 to 1 length-to-depth ratio and flaw depths of 1/8 and 1/4 of the base metal thickness. Various meshing schemes are evaluated for modelling the crack front profile, with particular emphasis on the region near the inside surface and at the base metal/cladding interface. The shape of the crack front profile through the cladding layer and the number of finite elements used to discretize the cladding thickness are found to have a significant influence on typical fracture mechanic measures of the crack tip stress fields. Results suggest that the stress intensity factor at the inner surface of a cladded vessel may be affected as much by the finite element mesh near the surface as by the material discontinuity between the two parts of the structure.« less

Stress corrosion cracking of Zircaloys in unirradiated and irradiated CsI

NASA Astrophysics Data System (ADS)

Cox, B.; Surette, B. A.; Wood, J. C.

1986-03-01

Unirradiated split-ring specimens of Zircaloy fuel cladding, coated with CsI, cracked when stressed at elevated temperatures. The specimens have been reexamined fractographically and metallographically in order to confirm that the cause of cracking was stress corrosion (SCC) and not delayed hydride cracking (DHC). Further specimens have been cracked at 350°C by a solution of CsI in a fused mixture of nitrates of rubidium, cesium, strontium and barium, by a similar mechanism. CsI dissolved in a fused molybdate melt was not stable at 400°C, and rapidly evolved iodine, leaving a melt that was incapable of causing SCC. Irradiation of stressed split-ring specimens of Zircaloy fuel cladding in a γ-irradiator of 10 6 R/h and in the U-5 loop in the NRU reactor at an estimated 10 9 R/h caused SCC when the specimens were packed in dry CsI powder. Care had to be taken to dry the CsI, otherwise cracking occurred by a DHC mechanism from hydrogen absorbed from residual moisture in the CsI. Fractography showed that the crack surfaces obtained with dry CsI were typical of iodine-induced SCC rather than cesium-induced metal vapour embrittlement. Thus, if a transport process is provided for the iodide to obtain access to the zirconium surface, CsI is capable of causing SCC of Zircaloy. This transport process might be ionic diffusion in a fission product oxide melt in the fuel-clad gap, however, radiolysis of CsI to form a volatile iodine species in a radiation field is the more probable explanation of PCI failures.

Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

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

Unal, Cetin; Galloway, Jack D.

2014-09-12

In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermalmore » swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.« less

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

2015-07-01

In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in these assessments are preliminary, and that additional data are necessary for these materials, most significantly under irradiation.

Phosphate-core silica-clad Er/Yb-doped optical fiber and cladding pumped laser.

PubMed

Egorova, O N; Semjonov, S L; Velmiskin, V V; Yatsenko, Yu P; Sverchkov, S E; Galagan, B I; Denker, B I; Dianov, E M

2014-04-07

We present a composite optical fiber with a Er/Yb co-doped phosphate-glass core in a silica glass cladding as well as cladding pumped laser. The fabrication process, optical properties, and lasing parameters are described. The slope efficiency under 980 nm cladding pumping reached 39% with respect to the absorbed pump power and 28% with respect to the coupled pump power. Due to high doping level of the phosphate core optimal length was several times shorter than that of silica core fibers.

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.

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Cost analysis of a growth guidance system compared with traditional and magnetically controlled growing rods for early-onset scoliosis: a US-based integrated health care delivery system perspective.

PubMed

Luhmann, Scott J; McAughey, Eoin M; Ackerman, Stacey J; Bumpass, David B; McCarthy, Richard E

2018-01-01

Treating early-onset scoliosis (EOS) with traditional growing rods (TGR) is effective but requires periodic surgical lengthening, risking complications. Alternatives include magnetically controlled growing rods (MCGR) that lengthen noninvasively and the growth guidance system (GGS), which obviate the need for active, distractive lengthenings. Previous studies have reported promising clinical effectiveness for GGS; however the direct medical costs of GGS compared to TGR and MCGR have not yet been explored. To estimate the cost of GGS compared with MCGR and TGR for EOS an economic model was developed from the perspective of a US integrated health care delivery system. Using dual-rod constructs, the model estimated the cumulative costs associated with initial implantation, rod lengthenings (TGR, MCGR), revisions due to device failure, surgical-site infections, device exchange, and final spinal fusion over a 6-year episode of care. Model parameters were from peer-reviewed, published literature. Medicare payments were used as a proxy for provider costs. Costs (2016 US$) were discounted 3% annually. Over a 6-year episode of care, GGS was associated with fewer invasive surgeries per patient than TGR (GGS: 3.4; TGR: 14.4) and lower cumulative costs than MCGR and TGR, saving $25,226 vs TGR. Sensitivity analyses showed that results were sensitive to changes in construct costs, rod breakage rates, months between lengthenings, and TGR lengthening setting of care. Within the model, GGS resulted in fewer invasive surgeries and deep surgical site infections than TGR, and lower cumulative costs per patient than both MCGR and TGR, over a 6-year episode of care. The analysis did not account for family disruption, pain, psychological distress, or compromised health-related quality of life associated with invasive TGR lengthenings, nor for potential patient anxiety surrounding the frequent MCGR lengthenings. Further analyses focusing strictly on current generation technologies should be considered for future research.

Cost analysis of a growth guidance system compared with traditional and magnetically controlled growing rods for early-onset scoliosis: a US-based integrated health care delivery system perspective

PubMed Central

Luhmann, Scott J; McAughey, Eoin M; Ackerman, Stacey J; Bumpass, David B; McCarthy, Richard E

2018-01-01

Purpose Treating early-onset scoliosis (EOS) with traditional growing rods (TGR) is effective but requires periodic surgical lengthening, risking complications. Alternatives include magnetically controlled growing rods (MCGR) that lengthen noninvasively and the growth guidance system (GGS), which obviate the need for active, distractive lengthenings. Previous studies have reported promising clinical effectiveness for GGS; however the direct medical costs of GGS compared to TGR and MCGR have not yet been explored. Methods To estimate the cost of GGS compared with MCGR and TGR for EOS an economic model was developed from the perspective of a US integrated health care delivery system. Using dual-rod constructs, the model estimated the cumulative costs associated with initial implantation, rod lengthenings (TGR, MCGR), revisions due to device failure, surgical-site infections, device exchange, and final spinal fusion over a 6-year episode of care. Model parameters were from peer-reviewed, published literature. Medicare payments were used as a proxy for provider costs. Costs (2016 US$) were discounted 3% annually. Results Over a 6-year episode of care, GGS was associated with fewer invasive surgeries per patient than TGR (GGS: 3.4; TGR: 14.4) and lower cumulative costs than MCGR and TGR, saving $25,226 vs TGR. Sensitivity analyses showed that results were sensitive to changes in construct costs, rod breakage rates, months between lengthenings, and TGR lengthening setting of care. Conclusion Within the model, GGS resulted in fewer invasive surgeries and deep surgical site infections than TGR, and lower cumulative costs per patient than both MCGR and TGR, over a 6-year episode of care. The analysis did not account for family disruption, pain, psychological distress, or compromised health-related quality of life associated with invasive TGR lengthenings, nor for potential patient anxiety surrounding the frequent MCGR lengthenings. Further analyses focusing strictly on current generation technologies should be considered for future research. PMID:29588607

Influence of Impact Damage on Carbon-Epoxy Stiffener Crippling

NASA Technical Reports Server (NTRS)

Jegley, Dawn C.

2010-01-01

NASA, the Air Force Research Laboratory and The Boeing Company have worked to develop new low-cost, light-weight composite structures for aircraft. A Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) concept has been developed which offers advantages over traditional metallic structure. In this concept a stitched carbon-epoxy material system has been developed with the potential for reducing the weight and cost of transport aircraft structure by eliminating fasteners, thereby reducing part count and labor. By adding unidirectional carbon rods to the top of stiffeners, the panel becomes more structurally efficient. This combination produces a more damage tolerant design. This document describes the results of experimentation on PRSEUS specimens loaded in unidirectional compression subjected to impact damage and loaded in fatigue and to failure. A comparison with analytical predictions for pristine and damaged specimens is included.

Hot Forging of a Cladded Component by Automated GMAW Process

NASA Astrophysics Data System (ADS)

Rafiq, Muhammad; Langlois, Laurent; Bigot, Régis

2011-01-01

Weld cladding is employed to improve the service life of engineering components by increasing corrosion and wear resistance and reducing the cost. The acceptable multi-bead cladding layer depends on single bead geometry. Hence, in first step, the relationship between input process parameters and the single bead geometry is studied and in second step a comprehensive study on multi bead clad layer deposition is carried out. This paper highlights an experimental study carried out to get single layer cladding deposited by automated Gas Metal Arc Welding (GMAW) process and to find the possibility of hot forming of the cladded work piece to get the final hot formed improved structure. GMAW is an arc welding process that uses an arc between a consumable electrode and the welding pool with an external shielding gas and the cladding is done by alongside deposition of weld beads. The experiments for single bead were conducted by varying the three main process parameters wire feed rate, arc voltage and welding speed while keeping other parameters like nozzle to work distance, shielding gas and its flow rate and torch angle constant. The effect of bead spacing and torch orientation on the cladding quality of single layer from the results of single bead deposition was studied. Effect of the dilution rate and nominal energy on the cladded layer hot bending quality was also performed at different temperatures.

Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

NASA Astrophysics Data System (ADS)

Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

2018-05-01

The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (< 1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase ( 10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He2+ implantation.

Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

NASA Astrophysics Data System (ADS)

Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

2018-04-01

The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (< 1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase ( 10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He2+ implantation.

Oxide dispersion strengthened ferritic steels: a basic research joint program in France

NASA Astrophysics Data System (ADS)

Boutard, J.-L.; Badjeck, V.; Barguet, L.; Barouh, C.; Bhattacharya, A.; Colignon, Y.; Hatzoglou, C.; Loyer-Prost, M.; Rouffié, A. L.; Sallez, N.; Salmon-Legagneur, H.; Schuler, T.

2014-12-01

AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9-14% Cr ferritic-martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical-chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α‧ unmixing.

Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

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

Katoh, Yutai; Terrani, Kurt A.

2015-08-01

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

Cyclic furnace oxidation of clad WI-52 systems at 1040 C and 1090 C

NASA Technical Reports Server (NTRS)

Gedwill, M. A.

1972-01-01

Cyclic furnace oxidation studies were conducted on the cobalt alloy WI-52 clad with Ni-30Cr, Fe-25Cr-4A1, and Ni-20Cr-4A1 foils (0.051 to 0.254 mm thick). Tests as long as 400 hours using 1- and 20-hour cycles showed that the Ni-Cr- and Fe-Cr-A1 claddings were about equally protective at both temperatures. The protective ability of these alloys was influenced by exposure temperature and cladding thickness. At both temperatures, they protected WI-52 about as well as, or better than, a widely used commercial aluminide coating. The Ni-Cr-Al claddings did not protect WI-52 nearly as well. Interdiffusion generally influenced the oxidation behavior of all clad WI-52 systems.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

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

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

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

Rectangular-cladding silicon slot waveguide with improved nonlinear performance

NASA Astrophysics Data System (ADS)

Huang, Zengzhi; Huang, Qingzhong; Wang, Yi; Xia, Jinsong

2018-04-01

Silicon slot waveguides have great potential in hybrid silicon integration to realize nonlinear optical applications. We propose a rectangular-cladding hybrid silicon slot waveguide. Simulation result shows that, with a rectangular-cladding, the slot waveguide can be formed by narrower silicon strips, so the two-photon absorption (TPA) loss in silicon is decreased. When the cladding material is a nonlinear polymer, the calculated TPA figure of merit (FOMTPA) is 4.4, close to the value of bulk nonlinear polymer of 5.0. This value confirms the good nonlinear performance of rectangular-cladding silicon slot waveguides.

Protective claddings for high strength chromium alloys

NASA Technical Reports Server (NTRS)

Collins, J. F.

1971-01-01

The application of a Cr-Y-Hf-Th alloy as a protective cladding for a high strength chromium alloy was investigated for its effectiveness in inhibiting nitrogen embrittlement of a core alloy. Cladding was accomplished by a combination of hot gas pressure bonding and roll cladding techniques. Based on bend DBTT, the cladding alloy was effective in inhibiting nitrogen embrittlement of the chromium core alloy for up to 720 ks (200hours) in air at 1422 K (2100 F). A significant increase in the bend DBTT occurred with longer time exposures at 1422 K or short time exposures at 1589 K (2400 F).

Orientation-Dependent Displacement Sensor Using an Inner Cladding Fiber Bragg Grating

PubMed Central

Yang, Tingting; Qiao, Xueguang; Rong, Qiangzhou; Bao, Weijia

2016-01-01

An orientation-dependent displacement sensor based on grating inscription over a fiber core and inner cladding has been demonstrated. The device comprises a short piece of multi-cladding fiber sandwiched between two standard single-mode fibers (SMFs). The grating structure is fabricated by a femtosecond laser side-illumination technique. Two well-defined resonances are achieved by the downstream both core and cladding fiber Bragg gratings (FBGs). The cladding resonance presents fiber bending dependence, together with a strong orientation dependence because of asymmetrical distribution of the “cladding” FBG along the fiber cross-section. PMID:27626427

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

Thermodynamic modelling of the C-U and B-U binary systems

NASA Astrophysics Data System (ADS)

Chevalier, P. Y.; Fischer, E.

2001-02-01

The thermodynamic modelling of the carbon-uranium (C-U) and boron-uranium (B-U) binary systems is being performed in the framework of the development of a thermodynamic database for nuclear materials, for increasing the basic knowledge of key phenomena which may occur in the event of a severe accident in a nuclear power plant. Applications are foreseen in the nuclear safety field to the physico-chemical interaction modelling, on the one hand the in-vessel core degradation producing the corium (fuel, zircaloy, steel, control rods) and on the other hand the ex-vessel molten corium-concrete interaction (MCCI). The key O-U-Zr ternary system, previously modelled, allows us to describe the first interaction of the fuel with zircaloy cladding. Then, the three binary systems Fe-U, Cr-U and Ni-U were modelled as a preliminary work for modelling the O-U-Zr-Fe-Cr-Ni multicomponent system, allowing us to introduce the steel components in the corium. In the existing database (TDBCR, thermodynamic data base for corium), Ag and In were introduced for modelling AIC (silver-indium-cadmium) control rods which are used in French pressurized water reactors (PWR). Elsewhere, B 4C is also used for control rods. That is why it was agreed to extend in the next years the database with two new components, B and C. Such a work needs the thermodynamic modelling of all the binary and pseudo-binary sub-systems resulting from the combination of B, B 2O 3 and C with the major components of TDBCR, O-U-Zr-Fe-Cr-Ni-Ag-In-Ba-La-Ru-Sr-Al-Ca-Mg-Si + Ar-H. The critical assessment of the very numerous experimental information available for the C-U and B-U binary systems was performed by using a classical optimization procedure and the Scientific Group Thermodata Europe (SGTE). New optimized Gibbs energy parameters are given, and comparisons between calculated and experimental equilibrium phase diagrams or thermodynamic properties are presented. The self-consistency obtained is quite satisfactory.

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

DOE PAGES

George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; ...

2014-09-29

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in themore » fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more 239Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Furthermore, implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235U enrichment and the additional UO 2 pellet volume enabled by using thinner cladding.« less

78 FR 40200 - Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-03

... breaches.'' Zircaloy is a type of zirconium alloy which includes both Zircaloy-2 and Zircaloy-4 cladding, but does not include M5 cladding. The M5 is a different type of zirconium alloy, which does not... ``zirconium alloy'' clad spent fuel assemblies in the 24PHB DSC, which would include both the ``zircaloy clad...

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

DOE PAGES

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; ...

2015-09-03

Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested inmore » INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less

Residual stress and crack initiation in laser clad composite layer with Co-based alloy and WC + NiCr

NASA Astrophysics Data System (ADS)

Lee, Changmin; Park, Hyungkwon; Yoo, Jaehong; Lee, Changhee; Woo, WanChuck; Park, Sunhong

2015-08-01

Although laser cladding process has been widely used to improve the wear and corrosion resistance, there are unwanted cracking issues during and/or after laser cladding. This study investigates the tendency of Co-based WC + NiCr composite layers to cracking during the laser cladding process. Residual stress distributions of the specimen are measured using neutron diffraction and elucidate the correlation between the residual stress and the cracking in three types of cylindrical specimens; (i) no cladding substrate only, (ii) cladding with 100% stellite#6, and (iii) cladding with 55% stellite#6 and 45% technolase40s. The microstructure of the clad layer was composed of Co-based dendrite and brittle eutectic phases at the dendritic boundaries. And WC particles were distributed on the matrix forming intermediate composition region by partial melting of the surface of particles. The overlaid specimen exhibited tensile residual stress, which was accumulated through the beads due to contraction of the coating layer generated by rapid solidification, while the non-clad specimen showed compressive. Also, the specimen overlaid with 55 wt% stellite#6 and 45 wt% technolase40s showed a tensile stress higher than the specimen overlaid with 100% stellite#6 possibly, due to the difference between thermal expansion coefficients of the matrix and WC particles. Such tensile stresses can be potential driving force to provide an easy crack path ways for large brittle fractures combined with the crack initiation sites such as the fractured WC particles, pores and solidification cracks. WC particles directly caused clad cracks by particle fracture under the tensile stress. The pores and solidification cracks also affected as initiation sites and provided an easy crack path ways for large brittle fractures.

Laser performance and modeling of RE3+:YAG double-clad crystalline fiber waveguides

NASA Astrophysics Data System (ADS)

Li, Da; Lee, Huai-Chuan; Meissner, Stephanie K.; Meissner, Helmuth E.

2018-02-01

We report on laser performance of ceramic Yb:YAG and single crystal Tm:YAG double-clad crystalline fiber waveguide (CFW) lasers towards the goal of demonstrating the design and manufacturing strategy of scaling to high output power. The laser component is a double-clad CFW, with RE3+:YAG (RE = Yb, Tm respectively) core, un-doped YAG inner cladding, and ceramic spinel or sapphire outer cladding. Laser performance of the CFW has been demonstrated with 53.6% slope efficiency and 27.5-W stable output power at 1030-nm for Yb:YAG CFW, and 31.6% slope efficiency and 46.7-W stable output power at 2019-nm for Tm:YAG CFW, respectively. Adhesive-Free Bond (AFB®) technology enables a designable refractive index difference between core and inner cladding, and designable core and inner cladding sizes, which are essential for single transverse mode CFW propagation. To guide further development of CFW designs, we present thermal modeling, power scaling and design of single transverse mode operation of double-clad CFWs and redefine the single-mode operation criterion for the double-clad structure design. The power scaling modeling of double-clad CFW shows that in order to achieve the maximum possible output power limited by the physical properties, including diode brightness, thermal lens effect, and simulated Brillion scattering, the length of waveguide is in the range of 0.5 2 meters. The length of an individual CFW is limited by single crystal growth and doping uniformity to about 100 to 200 mm lengths, and also by availability of starting crystals and manufacturing complexity. To overcome the limitation of CFW lengths, end-to-end proximity-coupling of CFWs is introduced.

Restrictive allograft syndrome (RAS): a novel form of chronic lung allograft dysfunction.

PubMed

Sato, Masaaki; Waddell, Thomas K; Wagnetz, Ute; Roberts, Heidi C; Hwang, David M; Haroon, Ayesha; Wagnetz, Dirk; Chaparro, Cecilia; Singer, Lianne G; Hutcheon, Michael A; Keshavjee, Shaf

2011-07-01

Bronchiolitis obliterans syndrome (BOS) with small-airway pathology and obstructive pulmonary physiology may not be the only form of chronic lung allograft dysfunction (CLAD) after lung transplantation. Characteristics of a form of CLAD consisting of restrictive functional changes involving peripheral lung pathology were investigated. Patients who received bilateral lung transplantation from 1996 to 2009 were retrospectively analyzed. Baseline pulmonary function was taken as the time of peak forced expiratory volume in 1 second (FEV(1)). CLAD was defined as irreversible decline in FEV(1) < 80% baseline. The most accurate threshold to predict irreversible decline in total lung capacity and thus restrictive functional change was at 90% baseline. Restrictive allograft syndrome (RAS) was defined as CLAD meeting this threshold. BOS was defined as CLAD without RAS. To estimate the effect on survival, Cox proportional hazards models and Kaplan-Meier analyses were used. Among 468 patients, CLAD developed in 156; of those, 47 (30%) showed the RAS phenotype. Compared with the 109 BOS patients, RAS patients showed significant computed tomography findings of interstitial lung disease (p < 0.0001). Prevalence of RAS was approximately 25% to 35% of all CLAD over time. Patient survival of RAS was significantly worse than BOS after CLAD onset (median survival, 541 vs 1,421 days; p = 0.0003). The RAS phenotype was the most significant risk factor of death among other variables after CLAD onset (hazard ratio, 1.60; confidential interval, 1.23-2.07). RAS is a novel form of CLAD that exhibits characteristics of peripheral lung fibrosis and significantly affects survival of lung transplant patients. Copyright © 2011 International Society for Heart and Lung Transplantation. Published by Elsevier Inc. All rights reserved.

Polarization characteristics of double-clad elliptical fibers.

PubMed

Zhang, F; Lit, J W

1990-12-20

A scalar variational analysis based on a Gaussian approximation of the fundamental mode of a double-clad elliptical fiber with a depressed inner cladding is studied. The polarization properties and graphic results are presented; they are given in terms of three parameters: the ratio of the major axis to the minor axis of the core, the ratio of the inner cladding major axis to the core major axis, and the difference between the core index and the inner cladding index. The variations of both the spot size and the field intensity with core ellipticity are examined. It is shown that high birefringence and dispersion-free orthogonal polarization modes can be obtained within the single-mode region and that the field intensity distribution may be more confined to the fiber center than in a single-clad elliptical fiber.

Nuclear fuel element

DOEpatents

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

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

An empirical-statistical model for laser cladding of Ti-6Al-4V powder on Ti-6Al-4V substrate

NASA Astrophysics Data System (ADS)

Nabhani, Mohammad; Razavi, Reza Shoja; Barekat, Masoud

2018-03-01

In this article, Ti-6Al-4V powder alloy was directly deposited on Ti-6Al-4V substrate using laser cladding process. In this process, some key parameters such as laser power (P), laser scanning rate (V) and powder feeding rate (F) play important roles. Using linear regression analysis, this paper develops the empirical-statistical relation between these key parameters and geometrical characteristics of single clad tracks (i.e. clad height, clad width, penetration depth, wetting angle, and dilution) as a combined parameter (PαVβFγ). The results indicated that the clad width linearly depended on PV-1/3 and powder feeding rate had no effect on it. The dilution controlled by a combined parameter as VF-1/2 and laser power was a dispensable factor. However, laser power was the dominant factor for the clad height, penetration depth, and wetting angle so that they were proportional to PV-1F1/4, PVF-1/8, and P3/4V-1F-1/4, respectively. Based on the results of correlation coefficient (R > 0.9) and analysis of residuals, it was confirmed that these empirical-statistical relations were in good agreement with the measured values of single clad tracks. Finally, these relations led to the design of a processing map that can predict the geometrical characteristics of the single clad tracks based on the key parameters.

VERA and VERA-EDU 3.5 Release Notes

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

Sieger, Matt; Salko, Robert K.; Kochunas, Brendan M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems.more » For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.« less

Clad fiber capacitor and method of making same

DOEpatents

Tuncer, Enis

2013-11-26

A clad capacitor and method of manufacture includes assembling a preform comprising a ductile, electrically conductive fiber; a ductile, electrically insulating cladding positioned on the fiber; a ductile, electrically conductive sleeve positioned over the cladding. One or more of the preforms are then bundled, heated and drawn along a longitudinal axis to decrease the diameter of the ductile components of the preform and fuse the preform into a unitized strand.

Experimental Study on Composite Light-weight Microporous Concrete Cladding Panels

NASA Astrophysics Data System (ADS)

Lida, Tian; Dongyan, Wang; Kang, Liu

2018-03-01

A new type of composite light-weight microporous concrete cladding panel was developed, with the compound function of retaining and heat preservation. Two specimens of the new cladding panel and connection detailing were made for out-of-plane bending experiment. The results indicate that the new cladding panel and its connection detailing are of sufficient stiffness, bearing capacity and deformability under wind load and out-of-plane seismic action.

Clad fiber capacitor and method of making same

DOEpatents

Tuncer, Enis

2012-12-11

A clad capacitor and method of manufacture includes assembling a preform comprising a ductile, electrically conductive fiber; a ductile, electrically insulating cladding positioned on the fiber; and a ductile, electrically conductive sleeve positioned over the cladding. One or more preforms are then bundled, heated and drawn along a longitudinal axis to decrease the diameter of the ductile components of the preform and fuse the preform into a unitized strand.

Effect of Temperature and Sheet Temper on Isothermal Solidification Kinetics in Clad Aluminum Brazing Sheet

NASA Astrophysics Data System (ADS)

Benoit, Michael J.; Whitney, Mark A.; Wells, Mary A.; Winkler, Sooky

2016-09-01

Isothermal solidification (IS) is a phenomenon observed in clad aluminum brazing sheets, wherein the amount of liquid clad metal is reduced by penetration of the liquid clad into the core. The objective of the current investigation is to quantify the rate of IS through the use of a previously derived parameter, the Interface Rate Constant (IRC). The effect of peak temperature and initial sheet temper on IS kinetics were investigated. The results demonstrated that IS is due to the diffusion of silicon (Si) from the liquid clad layer into the solid core. Reduced amounts of liquid clad at long liquid duration times, a roughened sheet surface, and differences in resolidified clad layer morphology between sheet tempers were observed. Increased IS kinetics were predicted at higher temperatures by an IRC model as well as by experimentally determined IRC values; however, the magnitudes of these values are not in good agreement due to deficiencies in the model when applied to alloys. IS kinetics were found to be higher for sheets in the fully annealed condition when compared with work-hardened sheets, due to the influence of core grain boundaries providing high diffusivity pathways for Si diffusion, resulting in more rapid liquid clad penetration.

Laser Cladding of Ultra-Thin Nickel-Based Superalloy Sheets.

PubMed

Gabriel, Tobias; Rommel, Daniel; Scherm, Florian; Gorywoda, Marek; Glatzel, Uwe

2017-03-10

Laser cladding is a well-established process to apply coatings on metals. However, on substrates considerably thinner than 1 mm it is only rarely described in the literature. In this work 200 µm thin sheets of nickel-based superalloy 718 are coated with a powder of a cobalt-based alloy, Co-28Cr-9W-1.5Si, by laser cladding. The process window is very narrow, therefore, a precisely controlled Yb fiber laser was used. To minimize the input of energy into the substrate, lines were deposited by setting single overlapping points. In a design of experiments (DoE) study, the process parameters of laser power, laser spot area, step size, exposure time, and solidification time were varied and optimized by examining the clad width, weld penetration, and alloying depth. The microstructure of the samples was investigated by optical microscope (OM) and scanning electron microscopy (SEM), combined with electron backscatter diffraction (EBSD) and energy dispersive X-ray spectroscopy (EDX). Similarly to laser cladding of thicker substrates, the laser power shows the highest influence on the resulting clad. With a higher laser power, the clad width and alloying depth increase, and with a larger laser spot area the weld penetration decreases. If the process parameters are controlled precisely, laser cladding of such thin sheets is manageable.

Laser Cladding of Ultra-Thin Nickel-Based Superalloy Sheets

PubMed Central

Gabriel, Tobias; Rommel, Daniel; Scherm, Florian; Gorywoda, Marek; Glatzel, Uwe

2017-01-01

Laser cladding is a well-established process to apply coatings on metals. However, on substrates considerably thinner than 1 mm it is only rarely described in the literature. In this work 200 µm thin sheets of nickel-based superalloy 718 are coated with a powder of a cobalt-based alloy, Co–28Cr–9W–1.5Si, by laser cladding. The process window is very narrow, therefore, a precisely controlled Yb fiber laser was used. To minimize the input of energy into the substrate, lines were deposited by setting single overlapping points. In a design of experiments (DoE) study, the process parameters of laser power, laser spot area, step size, exposure time, and solidification time were varied and optimized by examining the clad width, weld penetration, and alloying depth. The microstructure of the samples was investigated by optical microscope (OM) and scanning electron microscopy (SEM), combined with electron backscatter diffraction (EBSD) and energy dispersive X-ray spectroscopy (EDX). Similarly to laser cladding of thicker substrates, the laser power shows the highest influence on the resulting clad. With a higher laser power, the clad width and alloying depth increase, and with a larger laser spot area the weld penetration decreases. If the process parameters are controlled precisely, laser cladding of such thin sheets is manageable. PMID:28772639

Progress and Challenges of Ultrasonic Testing for Stress in Remanufacturing Laser Cladding Coating

PubMed Central

Yan, Xiao-Ling; Dong, Shi-Yun; Xu, Bin-Shi; Cao, Yong

2018-01-01

Stress in laser cladding coating is an important factor affecting the safe operation of remanufacturing components. Ultrasonic testing has become a popular approach in the nondestructive evaluation of stress, because it has the advantages of safety, nondestructiveness, and online detection. This paper provides a review of ultrasonic testing for stress in remanufacturing laser cladding coating. It summarizes the recent research outcomes on ultrasonic testing for stress, and analyzes the mechanism of ultrasonic testing for stress. Remanufacturing laser cladding coating shows typical anisotropic behaviors. The ultrasonic testing signal in laser cladding coating is influenced by many complex factors, such as microstructure, defect, temperature, and surface roughness, among others. At present, ultrasonic testing for stress in laser cladding coating can only be done roughly. This paper discusses the active mechanism of micro/macro factors in the reliability of stress measurement, as well as the impact of stress measurement on the quality and safety of remanufacturing components. Based on the discussion, this paper proposes strategies to nondestructively, rapidly, and accurately measure stress in remanufacturing laser cladding coating. PMID:29438309

Progress and Challenges of Ultrasonic Testing for Stress in Remanufacturing Laser Cladding Coating.

PubMed

Yan, Xiao-Ling; Dong, Shi-Yun; Xu, Bin-Shi; Cao, Yong

2018-02-13

Stress in laser cladding coating is an important factor affecting the safe operation of remanufacturing components. Ultrasonic testing has become a popular approach in the nondestructive evaluation of stress, because it has the advantages of safety, nondestructiveness, and online detection. This paper provides a review of ultrasonic testing for stress in remanufacturing laser cladding coating. It summarizes the recent research outcomes on ultrasonic testing for stress, and analyzes the mechanism of ultrasonic testing for stress. Remanufacturing laser cladding coating shows typical anisotropic behaviors. The ultrasonic testing signal in laser cladding coating is influenced by many complex factors, such as microstructure, defect, temperature, and surface roughness, among others. At present, ultrasonic testing for stress in laser cladding coating can only be done roughly. This paper discusses the active mechanism of micro/macro factors in the reliability of stress measurement, as well as the impact of stress measurement on the quality and safety of remanufacturing components. Based on the discussion, this paper proposes strategies to nondestructively, rapidly, and accurately measure stress in remanufacturing laser cladding coating.