zircaloy cladding tube: Topics by Science.gov

Sample records for zircaloy cladding tube

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

NASA Astrophysics Data System (ADS)

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

2008-12-01

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

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

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

Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the diametrical direction under the external pressure of the water environment, but potentially grow later in life in the diametrical direction once the expanding fuel pelletmore » contacts the cladding inner wall (Reference (b)). Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction (Reference (c)). Once above the solubility limit in Zircaloy, the hydride precipitates as zirconium hydride (References (c) through (j)). Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal (Reference (k)). Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface (References (l) through (n)), in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer. References (o) through (q) investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference (o), reported that the applied tensile stress increased corrosion rates above a critical stress level in 400 C and 475 C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed stress effect. Kim et al. (Reference (p)) and Kim and Kim (Reference (q)) more recently investigated the influence that an applied hoop stress has on the corrosion resistance of Zircaloy tubes in a 400 C steam and in a 350 C concentrated lithia water environment. Both of these studies found the applied tensile hoop stress to have no effect on cladding corrosion rates in the 400 C steam environment but to have accelerated corrosion in the lithiated water environment. In both cases, the corrosion acceleration in the lithiated water environment was attributed to the accumulation of the increased hydrogen picked up in the lithiated environment into the tensile regions of the test specimen. Dense hydride rims have been shown, independent of clad strain, to accelerate the corrosion of Zirconium alloys (References (r) and (s)), suggesting that the primary effect of applied stresses on the corrosion of Zircaloy in the above studies is through the accumulation of hydrogen at the oxide-to-metal interface and not through a direct mechanical breakdown of the passivating boundary layer. To further investigate the potential role of in-situ clad straining (or stress) on Zircaloy corrosion rates, two experimental studies were performed. First, several samples that were irradiated with and without an applied stress were destructively examined for the extent of corrosion occurring in strained and nonstrained regions of the test samples. The extent of corrosion was determined, posttest, by metallographic examination. Second, the corrosion process was monitored in-situ using electrochemical impedance spectroscopy on samples exposed out-of-reactor with and without an applied stress. Post test, these autoclave samples were also metallographically examined.« less
Effects of pretreatment processes for Zr electrorefining of oxidized Zircaloy-4 cladding tubes

NASA Astrophysics Data System (ADS)

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

2014-06-01

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

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

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

1982-01-01

Seven batches of Zircaloy-2 nuclear fuel cladding tubes with different textures were processed from tube shells of the same size, by different reduction routes, using pilger and 3-roll mills. Based on the texture data of these tubes, the texture control of Zircaloy tubing, the texture gradient across the wall, and the texture change during annealing were studied. The deformation texture of Zicaloy-2 tubing was dependent on the tool's curvature and was independent of the dimensions of the mother tubes. The different slopes of texture gradients were observed between the tubing of higher strain ration and that of lower strain ratio.
Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

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

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

2015-01-01

Neutron radiography for hydrogen analysis was performed with several Zircaloy-4 cladding samples with controlled hydrogen concentrations up to 1100 ppm. Hydrogen charging was performed in a process tube that was heated to facilitate hydrogen absorption by the metal. A correlation between the hydrogen concentration in the hydrided tubes and the neutron intensity was established, by which hydrogen content can be determined precisely in a small area (55 m x 55 m). Radiography analysis was also performed to evaluate the heating rate and its correlation with the hydrogen distribution through hydrided materials. In addition to radiography analysis, tomography experiments were performedmore » on Zircaloy-4 tube samples to study the local hydrogen distribution. Through tomography analysis a 3D reconstruction of the tube was evaluated in which an uneven hydrogen distribution in the circumferential direction can be observed.« less
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.
Surface treatment to form a dispersed Y2O3 layer on Zircaloy-4 tubes

NASA Astrophysics Data System (ADS)

Jung, Yang-Il; Kim, Hyun-Gil; Guim, Hwan-Uk; Lim, Yoon-Soo; Park, Jung-Hwan; Park, Dong-Jun; Yang, Jae-Ho

2018-01-01

Zircaloy-4 is a traditional zirconium-based alloy developed for application in nuclear fuel cladding tubes. The surfaces of Zircaloy-4 tubes were treated using a laser beam to increase their mechanical strength. Laser beam scanning of a tube coated with yttrium oxide (Y2O3) resulted in the formation of a dispersed oxide layer in the tube's surface region. Y2O3 particles penetrated the Zircaloy-4 during the laser treatment and were distributed uniformly in the surface region. The thickness of the dispersed oxide layer varied from 50 to 140 μm depending on the laser beam trajectory. The laser treatment also modified the texture of the tube. The preferred basal orientation along the normal to the tube surface disappeared, and a random structure appeared after laser processing. The most obvious result was an increase in the mechanical strength. The tensile strength of Zircaloy-4 increased by 10-20% with the formation of the dispersed oxide layer. The compressive yield stress also increased, by more than 15%. Brittle fracture was observed in the surface-treated samples during tensile and compressive deformation at room temperature; however, the fracture behavior was changed in ductile at elevated temperatures.
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
The influence of strain rate and hydrogen on the plane-strain ductility of Zircaloy cladding

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

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

1998-03-01

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

NASA Astrophysics Data System (ADS)

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

2017-05-01

The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study.
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.
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
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...
Estimation of ring tensile properties of steam oxidized Zircaloy-4 fuel cladding under simulated LOCA condition

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

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

Yan, Yong; Keiser, James R; Terrani, Kurt A

2014-01-01

Oxidation experiments were conducted at 1200 C in flowing steam with tubing specimens of Zircaloy-4, 317, 347 stainless steels, and the commercial FeCrAl alloy APMT. The purpose was to determine the oxidation behavior and post quench ductility of these alloys under postulated loss-of-coolant accident conditions. The parabolic rate constant for Zircaloy-4 tubing samples at 1200 were determined to be k = 2.173 107 g2/cm4/s C, in excellent agreement with the Cathcart-Pawel correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. The ductility of post quenched samples was evaluated by ring compression tests atmore » 135 C. For Zircaloy-4, the ductile to brittle transition occurs at an equivalent cladding reacted (ECR) of 19.3%. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR 50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to four hours.« less
Post-quench ductility evaluation of Zircaloy-4 and select iron alloys under design basis and extended LOCA conditions

NASA Astrophysics Data System (ADS)

Yan, Y.; Keiser, J. R.; Terrani, K. A.; Bell, G. L.; Snead, L. L.

2014-05-01

Oxidation experiments were conducted at 1200 °C in flowing steam with tubing specimens of Zircaloy-4, 317, 347 stainless steels, and the commercial FeCrAl alloy APMT. The purpose was to determine the oxidation behavior and post-quench ductility under postulated and extended LOCA conditions. The parabolic rate constant for Zircaloy-4 tubing samples at 1200 °C was determined to be k = 2.173 × 107 g2/cm4/s, in excellent agreement with the Cathcart-Pawel correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. The ductility of post-quenched samples was evaluated by ring compression tests at 135 °C. For Zircaloy-4, the ductile to brittle transition occurs at an equivalent cladding reacted (ECR) of 19.3%. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR ≈ 50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to 4 h.
Standard-less analysis of Zircaloy clad samples by an instrumental neutron activation method

NASA Astrophysics Data System (ADS)

Acharya, R.; Nair, A. G. C.; Reddy, A. V. R.; Goswami, A.

2004-03-01

A non-destructive method for analysis of irregular shape and size samples of Zircaloy has been developed using the recently standardized k0-based internal mono standard instrumental neutron activation analysis (INAA). The samples of Zircaloy-2 and -4 tubes, used as fuel cladding in Indian boiling water reactors (BWR) and pressurized heavy water reactors (PHWR), respectively, have been analyzed. Samples weighing in the range of a few tens of grams were irradiated in the thermal column of Apsara reactor to minimize neutron flux perturbations and high radiation dose. The method utilizes in situ relative detection efficiency using the γ-rays of selected activation products in the sample for overcoming γ-ray self-attenuation. Since the major and minor constituents (Zr, Sn, Fe, Cr and/or Ni) in these samples were amenable to NAA, the absolute concentrations of all the elements were determined using mass balance instead of using the concentration of the internal mono standard. Concentrations were also determined in a smaller size Zircaloy-4 sample by irradiating in the core position of the reactor to validate the present methodology. The results were compared with literature specifications and were found to be satisfactory. Values of sensitivities and detection limits have been evaluated for the elements analyzed.

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
Effects of hydrogen on thermal creep behaviour of Zircaloy fuel cladding

NASA Astrophysics Data System (ADS)

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

2018-01-01

Zirconium alloys are extensively used for nuclear fuel cladding. Creep is one of the most likely degradation mechanisms for fuel cladding during reactor operating and repository conditions. Fuel cladding tubes undergo waterside corrosion during service and hydrogen is produced as a result of it-a fraction of which is picked up by cladding. Hydrogen remains in solid solution up to terminal solid solubility and it precipitates as brittle hydride phase in the zirconium metal matrix beyond this limiting concentration. Hydrogen, either in solid solution or as precipitated hydride, alters the creep behaviour of zirconium alloys. The present article critically reviews the influence of hydrogen on thermal creep behaviour of zirconium alloys, develops the systematic understanding of this multifaceted phenomenon, and delineates the thrust areas which require further investigations.
Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

NASA Astrophysics Data System (ADS)

Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

2017-12-01

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.
Mechanistic Considerations Used in the Development of the PROFIT PCI Failure Model

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

Pankaskie, P. J.

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

NASA Astrophysics Data System (ADS)

Auzoux, Q.; Bouffioux, P.; Machiels, A.; Yagnik, S.; Bourdiliau, B.; Mallet, C.; Mozzani, N.; Colas, K.

2017-10-01

Precipitation of radial hydrides in zirconium-based alloy cladding concomitant with the cooling of spent nuclear fuel during dry storage can potentially compromise cladding integrity during its subsequent handling and transportation. This paper investigates hydride reorientation and its impact on ductility in unirradiated and irradiated recrystallized Zircaloy-2 cladding with an inner liner (cladding for boiling water reactors) subjected to hydride reorientation treatments. Cooling from 400 °C, hydride reorientation occurs in recrystallized Zircaloy-2 with liner at a lower effective stress in irradiated samples (below 40 MPa) than in unirradiated specimens (between 40 and 80 MPa). Despite significant hydride reorientation, unirradiated recrystallized Zircaloy-2 with liner cladding containing ∼200 wppm hydrogen shows a high diametral strain at fracture (>15%) during burst tests at ambient temperature. This ductile behavior is due to (1) the lower yield stress of the recrystallized cladding materials in comparison to hydride fracture strength (corrected by the compression stress arising from the precipitation) and (2) the hydride or hydrogen-depleted zone as a result of segregation of hydrogen into the liner layer. In irradiated Zircaloy-2 with liner cladding containing ∼340 wppm hydrogen, the conservation of some ductility during ring tensile tests at ambient temperature after reorientation treatment at 400 °C with cooling rates of ∼60 °C/h is also attributed to the existence of a hydride-depleted zone. Treatments at lower cooling rates (∼6 °C/h and 0.6 °C/h) promote greater levels of hydrogen segregation into the liner and allow for increased irradiation defect annealing, both of which result in a significant increase in ductility. Based on this investigation, given the very low cooling rates typical of dry storage systems, it can be concluded that the thermal transients associated with dry storage should not degrade, and more likely should actually improve, ductility of recrystallized Zircaloy-2 cladding with inner liner with such hydrogen content.
Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

DOE PAGES

Yan, Y.; Qian, S.; Littrell, K.; ...

2015-02-13

A non-destructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentration were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distributionmore » of circumferential hydrides across the wall. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. This study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount ( 20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor will be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for absolute hydrogen concentration determination.« less
SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

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

IJ van Rooyen; WR Lloyd; TL Trowbridge

2013-09-01

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

NASA Astrophysics Data System (ADS)

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

1993-09-01

The excellent thermal, chemical and neutronic properties of metallic fuel (U-Pu-Zr alloy) will lead to drastic improvements in fast reactor safety and the related fuel cycle economy. Some new high molybdenum 12Cr ferritic stainless steel candidate cladding alloys have been designed to achieve the mechanical properties required for high performance metallic fuel elements. These candidate claddings were irradiated by ion bombardment and tested to determine their strength and creep rupture properties. A 12Cr-8Mo and a 12Cr-8Mo-0.1Y 2O 3 steel were fabricated into cladding via a powder metallurgy process and by a mechanical alloying process, respectively. These claddings had two and three times the creep rupture strength (pressurized at 650°C for 10000 h) of a conventional 12Cr ferritic steel (HT-9). These two steels also showed no void formation up to 350 dpa by Ni 3+ irradiation. A zircaloy-2 lined steel cladding tube has also been fabricated for the purpose of reducing fuel-cladding interdiffusion and chemical interaction.
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 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.
Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering

NASA Astrophysics Data System (ADS)

Yan, Yong; Qian, Shuo; Garrison, Ben; Smith, Tyler; Kim, Peter

2018-04-01

A nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0 wt. % at 1100 °C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness, and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.
Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering

DOE PAGES

Yan, Yong; Qian, Shuo; Garrison, Ben; ...

2018-04-15

In this study, a nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0wt. % at 1100°C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness,more » and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.« less
Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering

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

Yan, Yong; Qian, Shuo; Garrison, Ben

In this study, a nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0wt. % at 1100°C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness,more » and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.« 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.
Recycle of Zirconium from Used Nuclear Fuel Cladding: A Major Element of Waste Reduction

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

Collins, Emory D; DelCul, Guillermo D; Terekhov, Dmitri

2011-01-01

Feasibility tests were initiated to determine if the zirconium in commercial used nuclear fuel (UNF) cladding can be recovered in sufficient purity to permit re-use, and if the recovery process can be operated economically. Initial tests are being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Early results indicate that quantitative recovery can be accomplished and product contamination with alloy constituents can be controlled sufficiently to meet purification requirements. Future tests with actual radioactive UNF claddingmore » are planned. The objective of current research is to determine the feasibility of recovery and recycle of zirconium from used fuel cladding wastes. Zircaloy cladding, which contains 98+% of hafnium-free zirconium, is the second largest mass, on average {approx}25 wt %, of the components in used U.S. light-water-reactor fuel assemblies. Therefore, recovery and recycle of the zirconium would enable a large reduction in geologic waste disposal for advanced fuel cycles. Current practice is to compact or grout the cladding waste and store it for subsequent disposal in a geologic repository. This paper describes results of initial tests being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Future tests with actual radioactive UNF cladding are planned.« less
On the corrosion behavior of zircaloy-4 in spent fuel pools under accidental conditions

NASA Astrophysics Data System (ADS)

Lavigne, O.; Shoji, T.; Sakaguchi, K.

2012-07-01

After zircaloy cladding tubes have been subjected to irradiation in the reactor core, they are stored temporarily in spent fuel pools. In case of an accident, the integrity of the pool may be affected and the composition of the coolant may change drastically. This was the case in Fukushima Daiichi in March 2011. Successive incidents have led to an increase in the pH of the coolant and to chloride contamination. Moreover, water radiolysis may occur owing to the remnant radioactivity of the spent fuel. In this study, we propose to evaluate the corrosion behavior of oxidized Zr-4 (in autoclave at 288 °C for 32 days) in function of the pH and the presence of chloride and radical forms. The generation of radicals is achieved by the sonolysis of the solution. It appears that the increase in pH and the presence of radicals lead to an increase in current densities. However, the current densities remain quite low (depending on the conditions, between 1 and 10 μA cm-2). The critical parameter is the presence of chloride ions. The chloride ions widely decrease the passive range of the oxidized samples (the pitting potential is measured around +0.6 V (vs. SCE)). Moreover, if the oxide layer is scratched or damaged (which is likely under accidental conditions), the pitting potential of the oxidized sample reaches the pitting potential of the non-oxidized sample (around +0.16 V (vs. SCE)), leaving a shorter stable passive range for the Zr-4 cladding tubes.
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
A novel procedure employing laser ultrasound technique and simplex algorism for the characterization of mechanical and geometrical properties in Zircaloy tubes with different levels of hydrogen charging

NASA Astrophysics Data System (ADS)

Liu, I.-Hung; Yang, Che-Hua

2011-01-01

In this research, a procedure employing a laser ultrasound technique (LUT) and an inversion algorism is reported for nondestructive characterization of mechanical and geometrical properties in Zircaloy tubes with different levels of hydrogen charging. With the LUT, guided acoustic waves are generated to propagate in the Zircaloy tubes and are detected remotely by optical means. By measuring the dispersive wavespeeds followed by the inversion algorism, mechanical properties such as elastic moduli and geometrical property such as wall-thickness of Zircaloy tubes are characterized for different levels of hydrogen charging. Having the advantages of remote, non-contact and point-wise generation/detection, the reported procedure serves as a competitive candidate for the characterization of Zircaloy tubes generally operated in irradiative and temperature-elevated environments.
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

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

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

Terrani, Kurt A.; Petrie, Christian M.

2016-05-01

This calculation summarizes thermal analyses of a flexible rabbit design for irradiating a variety of pressurized water reactor (PWR) cladding materials (stainless steel, iron-chromium aluminum [FeCrAl], Zircaloy, and Inconel) with variable dimensions at a temperature of 350 °C in the flux trap of the High Flux Isotope Reactor (HFIR). The design can accommodate standard cladding for outer diameters (ODs) of approximately 9.50 mm with thickness ranging from 0.30 mm to 0.70 mm. The length is generally between 10 and 50 mm. The specimens contain moly inserts with a variable OD that provides the heat flux necessary to achieve the designmore » temperature with such a small fixed gas gap. The primary outer containment is an Al-6061 housing with a slightly enlarged inner diameter (ID) of 9.60 mm. The specimen temperature is controlled by determining a helium/argon gas mixture specific to the as-built specimen and housing. Variables that affect the required gas mixture are the cladding material (thermal expansion, density, heat generation rate), cladding OD, housing ID, and cladding ID. This calculation documents the analyses performed to determine required gas mixtures for a variety of scenarios.« less
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
High-temperature oxidation kinetics of sponge-based E110 cladding alloy

DOE PAGES

Yan, Yong; Garrison, Benton E.; Howell, Mike; ...

2017-11-03

Two-sided oxidation experiments were recently conducted at 900°C–1200 °C in flowing steam with samples of sponge-based Zr-1Nb alloy E110. Although the old electrolytic E110 tubing exhibited a high degree of susceptibility to nodular corrosion and experienced breakaway oxidation rates in a relatively short time, the new sponge-based E110 demonstrated steam oxidation behavior comparable to Zircaloy-4. Sample weight gain and oxide layer thickness measurements were performed on oxidized E110 specimens and compared to oxygen pickup and oxide layer thickness calculations using the Cathcart-Pawel correlation. Our study shows that the sponge-based E110 follows the parabolic law at temperatures above 1015 °C. Atmore » or below 1015 °C, the oxidation rate was very low when compared to Zircaloy-4 and can be represented by a cubic expression. No breakaway oxidation was observed at 1000 °C for oxidation times up to 10,000 s. Arrhenius expressions are given to describe the parabolic rate constants at temperatures above 1015 °C and cubic rate constants are provided for temperatures below 1015 °C. The weight gains calculated by our equations are in excellent agreement with the measured sample weight gains at all test temperatures. In addition to the as-fabricated E110 cladding sample, prehydrided E110 cladding with hydrogen concentrations in the 100–150 wppm range was also investigated. The effect of hydrogen content on sponge-based E110 oxidation kinetics was minimal. No significant difference was found between as-fabricated and hydrided samples with regard to oxygen pickup and oxide layer thickness for hydrogen contents below 150 wppm.« less
High-temperature oxidation kinetics of sponge-based E110 cladding alloy

NASA Astrophysics Data System (ADS)

Yan, Yong; Garrison, Benton E.; Howell, Mike; Bell, Gary L.

2018-02-01

Two-sided oxidation experiments were recently conducted at 900°C-1200 °C in flowing steam with samples of sponge-based Zr-1Nb alloy E110. Although the old electrolytic E110 tubing exhibited a high degree of susceptibility to nodular corrosion and experienced breakaway oxidation rates in a relatively short time, the new sponge-based E110 demonstrated steam oxidation behavior comparable to Zircaloy-4. Sample weight gain and oxide layer thickness measurements were performed on oxidized E110 specimens and compared to oxygen pickup and oxide layer thickness calculations using the Cathcart-Pawel correlation. Our study shows that the sponge-based E110 follows the parabolic law at temperatures above 1015 °C. At or below 1015 °C, the oxidation rate was very low when compared to Zircaloy-4 and can be represented by a cubic expression. No breakaway oxidation was observed at 1000 °C for oxidation times up to 10,000 s. Arrhenius expressions are given to describe the parabolic rate constants at temperatures above 1015 °C and cubic rate constants are provided for temperatures below 1015 °C. The weight gains calculated by our equations are in excellent agreement with the measured sample weight gains at all test temperatures. In addition to the as-fabricated E110 cladding sample, prehydrided E110 cladding with hydrogen concentrations in the 100-150 wppm range was also investigated. The effect of hydrogen content on sponge-based E110 oxidation kinetics was minimal. No significant difference was found between as-fabricated and hydrided samples with regard to oxygen pickup and oxide layer thickness for hydrogen contents below 150 wppm.
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
Study of iodine migration in zirconia using stable and radioactive ion implantation

NASA Astrophysics Data System (ADS)

Chevarier, N.; Brossard, F.; Chevarier, A.; Crusset, D.; Moncoffre, N.

1998-03-01

The large uranium fission cross section leading to iodine and the behaviour of this element in the cladding tube during energy production and afterwards during waste storage is a crucial problem, especially for 129I which is a very long half-life isotope ( T = 1.59 × 10 7yr). Since a combined external and internal oxidation of the zircaloy cladding tube occurs during the reactor processing, iodine diffusion parameters in zirconia are needed. In order to obtain these data, stable iodine atoms were first introduced by ion implantation into zirconia with an energy of 200 keV and a dose equal to 8 × 10 15at cm -2. Diffusion profiles were measured using 3 MeV alpha-particle Rutherford Backscattering Spectrometry at each step of the annealing procedure between 700°C and 900°C. In such experiments a reduced iodine concentration was observed, which correlated to a diffusion-like process. Similar analysis has been performed using radioactive 131I implanted at a very low dose of 10 9 at cm -2. In this case the iodine release is deduced from gamma-ray spectroscopy measurements. The results are discussed in this paper.
Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

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

Isabella J van Rooyen

2012-09-01

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrixmore » composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing« less
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
Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

NASA Astrophysics Data System (ADS)

Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

2017-04-01

A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.
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).
Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

NASA Astrophysics Data System (ADS)

Wang, Hong; Wang, Jy-An John

2016-10-01

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

DOE PAGES

Wang, Hong; Wang, Jy-An John

2016-07-20

We studied behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending. Tests were performed under load or moment control at 5 Hz, and an empirical correlation was established between rod fatigue life and amplitude of the applied moment. Fatigue response of Zry-4 cladding was further characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment applied and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition all affect surrogate rod failure. Bonding/debonding of PPI/PCI and pellet fracturing contribute to surrogatemore » rod bending fatigue. Also, the effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective specimen gauge length is effective in sensor spacing correction. Finally, we developed the database and gained understanding in this study such that it will serve as input to analysis of SNF vibration integrity.« less
75 FR 80546 - Virginia Electric and Power Company; Surry Power Station Unit Nos. 1 and 2; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-22

... used to predict the rates of energy release, hydrogen concentration, and cladding oxidation from the... associated hydrogen pickup) for Optimized ZIRLO TM at any given burnup would be less than both zircaloy-4 and... between cladding hydrogen content (due to in-service corrosion) and post-quench ductility. \\2\\ ADAMS...
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.
Hydrogen motion in Zircaloy-4 cladding during a LOCA transient

NASA Astrophysics Data System (ADS)

Elodie, T.; Jean, D.; Séverine, G.; M-Christine, B.; Michel, C.; Berger, P.; Martine, B.; Antoine, A.

2016-04-01

Hydrogen and oxygen are key elements influencing the embrittlement of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). The understanding of the mechanisms influencing the motion of these two chemical elements in the metal is required to fully describe the material embrittlement. High temperature steam oxidation tests were performed on pre-hydrided Zircaloy-4 samples with hydrogen contents ranging between 11 and 400 wppm prior to LOCA transient. Thanks to the use of both Electron Probe Micro-Analysis (EPMA) and Elastic Recoil Detection Analysis (μ-ERDA), the chemical elements partitioning has been systematically quantified inside the prior-β phase. Image analysis and metallographic examinations were combined to provide an average oxygen profile as well as hydrogen profile within the cladding thickness after LOCA transient. The measured hydrogen profile is far from homogeneous. Experimental distributions are compared to those predicted numerically using calculations derived from a finite difference thermo-diffusion code (DIFFOX) developed at IRSN.
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.
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

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.
Multispectral pyrometry for surface temperature measurement of oxidized Zircaloy claddings

NASA Astrophysics Data System (ADS)

Bouvry, B.; Cheymol, G.; Ramiandrisoa, L.; Javaudin, B.; Gallou, C.; Maskrot, H.; Horny, N.; Duvaut, T.; Destouches, C.; Ferry, L.; Gonnier, C.

2017-06-01

Non-contact temperature measurement in a nuclear reactor is still a huge challenge because of the numerous constraints to consider, such as the high temperature, the steam atmosphere, and irradiation. A device is currently developed at CEA to study the nuclear fuel claddings behavior during a Loss-of-Coolant Accident. As a first step of development, we designed and tested an optical pyrometry procedure to measure the surface temperature of nuclear fuel claddings without any contact, under air, in the temperature range 700-850 °C. The temperature of Zircaloy-4 cladding samples was retrieved at various temperature levels. We used Multispectral Radiation Thermometry with the hypothesis of a constant emissivity profile in the spectral ranges 1-1.3 μm and 1.45-1.6 μm. To allow for comparisons, a reference temperature was provided by a thermocouple welded on the cladding surface. Because of thermal losses induced by the presence of the thermocouple, a heat transfer simulation was also performed to estimate the bias. We found a good agreement between the pyrometry measurement and the temperature reference, validating the constant emissivity profile hypothesis used in the MRT estimation. The expanded measurement uncertainty (k = 2) of the temperature obtained by the pyrometry method was ±4 °C, for temperatures between 700 and 850 °C. Emissivity values, between 0.86 and 0.91 were obtained.
Accident-tolerant oxide fuel and cladding

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

Mariani, Robert D.

Systems and methods for accident tolerant oxide fuel. One or more disks can be placed between fuel pellets comprising UO.sub.2, wherein such disks possess a higher thermal conductivity material than that of the UO.sub.2 to provide enhanced heat rejection thereof. Additionally, a cladding coating comprising zircaloy coated with a material that provides stability and high melting capability can be provided. The pellets can be configured as annular pellets having an annulus filled with the higher thermal conductivity material. The material coating the zircaloy can be, for example, Zr.sub.5Si.sub.4 or another silicide such as, for example, a Zr-Silicide that limits corrosion.more » The aforementioned higher thermal conductivity material can be, for example, Si, Zr.sub.xSi.sub.y, Zr, or Al.sub.2O.sub.3.« less
Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

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

Isabella J van Rooyen

2012-09-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.
Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

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

Isabella J van Rooyen

2013-01-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.
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.
An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

DOE PAGES

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

2017-04-30

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

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

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

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

PubMed

Reynolds, Jacob G; Huber, Heinz J; Cooke, Gary A; Pestovich, John A

2014-08-15

The United States Department of Energy Hanford Site, near Richland, Washington, USA, processed plutonium between 1944 and 1987. Fifty-six million gallons of waste of various origins remain, including waste from removing zircaloy fuel cladding using the so-called Zirflex process. The speciation of zirconium and fluoride in this waste is important because of the corrosivity and reactivity of fluoride as well as the (potentially) high density of Zr-phases. This study evaluates the solid-phase speciation of zirconium and fluoride using X-ray diffraction (XRD) and scanning electron microscopy with energy dispersive spectroscopy (SEM-EDS). Two waste samples were analyzed: one waste sample that is relatively pure zirconium cladding waste from tank 241-AW-105 and another that is a blend of zirconium cladding wastes and other high-level wastes from tank 241-C-104. Villiaumite (NaF) was found to be the dominant fluoride species in the cladding waste and natrophosphate (Na7F[PO4]2 · 19H2O) was the dominant species in the blended waste. Most zirconium was present as a sub-micron amorphous Na-Zr-O phase in the cladding waste and a Na-Al-Zr-O phase in the blended waste. Some zirconium was present in both tanks as either rounded or elongated crystalline needles of Na-bearing ZrO2 that are up to 200 μm in length. These results provide waste process planners the speciation data needed to develop disposal processes for this waste. Copyright © 2014 Elsevier B.V. All rights reserved.
Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding

NASA Astrophysics Data System (ADS)

Chollet, Mélanie; Valance, Stéphane; Abolhassani, Sousan; Stein, Gene; Grolimund, Daniel; Martin, Matthias; Bertsch, Johannes

2017-05-01

For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO2 are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components.
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.
A study on the reaction of Zircaloy-4 tube with hydrogen/steam mixture

NASA Astrophysics Data System (ADS)

Lee, Ji-Min; Kook, Dong-Hak; Cho, Il-Je; Kim, Yong-Soo

2017-08-01

In order to fundamentally understand the secondary hydriding mechanism of zirconium alloy cladding, the reaction of commercial Zircaloy-4 tubes with hydrogen and steam mixture was studied using a thermo-gravimetric analyser with two variables, H2/H2O ratio and temperature. Phenomenological analysis revealed that in the steam starvation condition, i.e., when the H2/H2O ratio is greater than 104, hydriding is the dominant reaction and the weight gain increases linearly after a short incubation time. On the other hand, when the gas ratio is 5 × 102 or 103, both hydriding and oxidation reactions take place simultaneously, leading to three distinct regimes: primary hydriding, enhanced oxidation, and massive hydriding. Microstructural changes of oxide demonstrate that when the weight gain exceeds a certain critical value, massive hydriding takes place due to the significant localized crack development within the oxide, which possibly simulates the secondary hydriding failure in a defective fuel operation. This study reveals that the steam starvation condition above the critical H2/H2O ratio is only a necessary condition for the secondary hydriding failure and, as a sufficient condition, oxide needs to grow sufficiently to reach the critical thickness that produces substantial crack development. In other words, in a real defective fuel operation incident, the secondary failure is initiated only when both steam starvation and oxide degradation conditions are simultaneously met. Therefore, it is concluded that the indispensable time for the critical oxide growth primarily determines the triggering time of massive hydriding failure.
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.
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
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.
Characterization of deformation mechanisms in zirconium alloys: effect of temperature and irradiation

NASA Astrophysics Data System (ADS)

Long, Fei

Zirconium alloys have been widely used in the CANDU (CANada Deuterium Uranium) reactor as core structural materials. Alloy such as Zircaloy-2 has been used for calandria tubes; fuel cladding; the pressure tube is manufactured from alloy Zr-2.5Nb. During in-reactor service, these alloys are exposed to a high flux of fast neutron at elevated temperatures. It is important to understand the effect of temperature and irradiation on the deformation mechanism of zirconium alloys. Aiming to provide experimental guidance for future modeling predictions on the properties of zirconium alloys this thesis describes the result of an investigation of the change of slip and twinning modes in Zircaloy-2 and Zr-2.5Nb as a function of temperature and irradiation. The aim is to provide scientific fundamentals and experimental evidences for future industry modeling in processing technique design, and in-reactor property change prediction of zirconium components. In situ neutron diffraction mechanical tests carried out on alloy Zircaloy-2 at three temperatures: 100¢ªC, 300¢ªC, and 500¢ªC, and described in Chapter 3. The evolution of the lattice strain of individual grain families in the loading and Poisson's directions during deformation, which probes the operation of slip and twinning modes at different stress levels, are described. By using the same type of in situ neutron diffraction technique, tests on Zr-2.5Nb pressure tube material samples, in either the fast-neutron irradiated or un-irradiated condition, are reported in Chapter 4. In Chapter 5, the measurement of dislocation density by means of line profile analysis of neutron diffraction patterns, as well as TEM observations of the dislocation microstructural evolution, is described. In Chapter 6 a hot-rolled Zr-2.5Nb with a larger grain size compared with the pressure tubing was used to study the development of dislocation microstructures with increasing plastic strain. In Chapter 7, in situ loading of heavy ion irradiated hot-rolled Zr-2.5Nb alloy is described, providing evidence for the interaction between moving dislocations and irradiation induced loops. Chapter 8 gives the effect on the dislocation structure of different levels of compressive strains along two directions in the hot-rolled Zr-2.5Nb alloy. By using high resolution neutron diffraction and TEM observations, the evolution of type and dislocation densities, as well as changes of dislocation microstructure with plastic strain were characterized.
(Project 13-5292) Correlating thermal and mechanical coupling based multiphysics behavior of nuclear materials through in-situ measurements

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

Tomar, Vikas

Irradiations and post characterization experiments were performed first on Zr samples. This step will help understand the effect of the 2.5% alloying elements on the behavior of Zircaloy-4 (PWR cladding material) when compared to pure Zr. Irradiation flux measurements and sample temperature calibrations were performed at different energies prior to the irradiation experiments. Irradiations were performed with two different energy regimes1: non-displacment energies and displacement energies. Time was also dedicated to optimize transmission electron microscopy (TEM) sample preparation conditions via electropolishing technique. This step is crucial to prepare TEM samples for the in-situ TEM/irradiation experiments (Year 2). In addition, Zircaloy-4more » samples are being prepared for irradiation, and a setup is built by one of our collaborators (Dr. Mert Efe) to prepare ultrafine (UF) and nanocrystalline (NC) Zircaloy-4 samples for comparison with the commercial Zircaloy-4 samples.« less
Hydride Microstructure at the Metal-Oxide Interface of Zircaloy-4 from H.B. Robinson Nuclear Reactor

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

Cinbiz, Mahmut N; Edmondson, Philip D; Terrani, Kurt A

2017-01-01

This study investigates the hydride rim microstructure at the metal-oxide interface of Zircaloy-4 cladding segment removed from H.B. Robinson Nuclear Reactor by utilizing high resolution electron microscopy techniques with energy dispersive x-ray spectroscopy at Oak Ridge National Laboratory under the NSUF Rapid Turnout Experiment program. A complex stacking and orientation of hydride platelets has been observed below the sub-oxide layer. Furthermore, radial hydride platelets have been observed. EDS signals of both Fe and Cr has been reduced within hydrides whereas EDS signal of Sn is unaffected.
HRTEM and chemical study of an ion-irradiated chromium/zircaloy-4 interface

NASA Astrophysics Data System (ADS)

Wu, A.; Ribis, J.; Brachet, J.-C.; Clouet, E.; Leprêtre, F.; Bordas, E.; Arnal, B.

2018-06-01

Chromium-coated zirconium alloys are being studied as Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors (LWRs). Those materials are especially studied to improve the oxidation resistance of LWRs current fuel claddings in nominal and at High Temperature (HT) for hypothetical accidental conditions such as LOss of Coolant Accident. Beyond their HT behavior, it is essential to assess the materials behavior under irradiation. A first generation chromium/Zircaloy-4 interface was thus irradiated with 20 MeV Kr8+ ions at 400 °C up to 10 dpa. High-Resolution Transmission Electron Microscopy and chemical analysis (EDS) were conducted at the Cr/Zr interface. The atomic structure of the interface reveals the presence of Zr(Fe, Cr)2 Laves phase, displaying both C14 and C15 structure. After irradiation, only the C14 structure was observed and atomic row matching was preserved across the different interfaces, thus ensuring a good adhesion of the coating after irradiation.

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
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.
75 FR 76051 - Northern States Power Company-Minnesota, Prairie Island Nuclear Generating Plant, Units 1 and 2...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-07

..., 2010 (Agencywide Documents Access and Management System Accession Nos. ML093280883 and ML101480083... systems for light-water nuclear power reactors,'' and appendix K to 10 CFR part 50, ``ECCS Evaluation... core cooling system (ECCS) for reactors fueled with zircaloy or ZIRLO\\TM\\ cladding. In addition...
DOE Office of Scientific and Technical Information (OSTI.GOV)

Lombardo, N.J.; Marseille, T.J.; White, M.D.

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic inmore » form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.« less
Evaluation of steam corrosion and water quenching behavior of zirconium-silicide coated LWR fuel claddings

NASA Astrophysics Data System (ADS)

Yeom, Hwasung; Lockhart, Cody; Mariani, Robert; Xu, Peng; Corradini, Michael; Sridharan, Kumar

2018-02-01

This study investigates steam corrosion of bulk ZrSi2, pure Si, and zirconium-silicide coatings as well as water quenching behavior of ZrSi2 coatings to evaluate its feasibility as a potential accident-tolerant fuel cladding coating material in light water nuclear reactor. The ZrSi2 coating and Zr2Si-ZrSi2 coating were deposited on Zircaloy-4 flats, SiC flats, and cylindrical Zircaloy-4 rodlets using magnetron sputter deposition. Bulk ZrSi2 and pure Si samples showed weight loss after the corrosion test in pure steam at 400 °C and 10.3 MPa for 72 h. Silicon depletion on the ZrSi2 surface during the steam test was related to the surface recession observed in the silicon samples. ZrSi2 coating (∼3.9 μm) pre-oxidized in 700 °C air prevented substrate oxidation but thin porous ZrO2 formed on the coating. The only condition which achieved complete silicon immobilization in the oxide scale in aqueous environments was the formation of ZrSiO4 via ZrSi2 coating oxidation in 1400 °C air. In addition, ZrSi2 coatings were beneficial in enhancing quenching heat transfer - the minimum film boiling temperature increased by 6-8% in the three different environmental conditions tested. During repeated thermal cycles (water quenching from 700 °C to 85 °C for 20 s) performed as a part of quench tests, no spallation and cracking was observed and the coating prevented oxidation of the underlying Zircaloy-4 substrate.
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.
Recovery and recrystalization kinetics of cold-worked Zircaloy-4 plate and tubing (LWBR Development Program)

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

Katz, O.M.

1968-02-01

Empirical kinetic equations were derived to describe the recovery region between 550 and 1020/sup 0/F for times to 4000 hours for 15 to 78% cold-worked Zircaloy-4 plate and tubing. The properties studied were electrical resistivity and X-ray line sharpening. Recrystallization kinetics were described with sigmoidal curves derived from X-ray intensity and microhardness data. Light, replica, and transmission electron microscopy and selected-area electron diffraction were used to postulate recovery and recrystallization mechanisms. From a structural aspect, the annealing process in cold-worked Zircaloy-4 is visualized as a dislocation climb and annihilation process to the limit allowed by the size of the deformationmore » subcells, a reorientation of the subgrain material into a recrystallization texture, a growth of reoriented cells located in the most highly worked bands, and a consumption of less favorably strained and/or oriented cells by the high-angle boundaries of the reoriented cells. Comparison of 15 and 73% cold-worked tubing showed the activation energy to be less (21 versus 60 kcal/mol) and the subcell size greater (8000A versus 1000A) for the 15% cold-worked material. (NSA 22: 21698)« less
Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

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

Koyanagi, Takaaki; Petrie, Christian M.

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

NASA Technical Reports Server (NTRS)

Kovach, C. W.

1978-01-01

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

DOEpatents

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

1984-01-01

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

DOEpatents

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

1985-01-01

An enlarged funnel is releasably mounted at the open end of a length of cladding by an encircling length of shrink tubing which securely engages outer surfaces of both the funnel and cladding. The shrink tubing overlaps an annular shoulder against which pulling force can be exerted to remove the tubing from the cladding. The shoulder can be provided on a separate collar or ring, or on the funnel itself.
Chemical and microstructural characterization of a 9 cycle Zircaloy-2 cladding using EPMA and FIB tomography

NASA Astrophysics Data System (ADS)

Baris, A.; Restani, R.; Grabherr, R.; Chiu, Y.-L.; Evans, H. E.; Ammon, K.; Limbäck, M.; Abolhassani, S.

2018-06-01

A high burn-up Zircaloy-2 cladding is characterised in order to correlate its microstructure and composition to the change of oxidation and hydrogen uptake behaviour during long term service in the reactor. After 9 cycle of service, the chemical analysis of the cladding segment shows that most secondary phase particles (SPPs) have dissolved into the matrix. Fe and Ni are distributed homogenously in the metal matrix. Cr-containing clusters, remnants of the original Zr(Fe, Cr)2 type precipitates, are still present. Hydrides are observed abundantly in the metal side close to the metal-oxide interface. These hydrides have lower Fe and Ni concentration than that in the metal matrix. The three-dimensional (3D) reconstruction of the oxide and the metal-oxide interface obtained by Focused Ion Beam (FIB) tomography shows how the oxide microstructure has evolved with the number of cycles. The composition and microstructural changes in the oxide and the metal can be correlated to the oxidation kinetics and the H-uptake. It is observed that there is an increase in the oxidation kinetics and in the H-uptake between the third and the fifth cycles, as well as during the last two cycles. At the same time the volume fraction of cracks in the oxide significantly increased. Many fine cracks and pores exist in the oxide formed in the last cycle. Furthermore, the EPMA results confirm that this oxide formed at the last cycle reflects the composition of the metal at the metal-oxide interface after the long residence time in the reactor.
Safety margins in zircaloy oxidation and embrittlement criteria for emergency core cooling system acceptance

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

Williford, R.E.

1986-09-01

Current emergency core cooling system acceptance criteria for light water reactors specify that, under loss-of-coolant accident (LOCA) conditions, the Baker-Just (BJ) correlation must be used to calculate Zircaloy-steam oxidation, calculated peak cladding temperatures (PCT) must not exceed 1204/sup 0/C, and calculated oxidation must not exceed 17% equivalent cladding reacted (ECR). An appropriately defined minimum margin of safety was estimated for each of these criteria. The currently required BJ oxidation correlation provides margins only over the 1100 to 1500/sup 0/C temperature range at the 95% confidence level. The PCT margins for thermal shock and handling failures are adequate at oxidation temperaturesmore » above 1204/sup 0/C for up to 210 and 160 s, respectively, at the 95% confidence level. The ECR thermal shock and handling margins at the 50 and 95% confidence levels, respectively, range between 2 and 7% ECR for the BJ correlation, but vanish at temperatures above 1100 to 1160/sup 0/C for the best-estimate Cathcart-Pawel correlation. However, use of the Cathcart Pawel correlation for ''design basis'' LOCA calculations can be justified at the 85 to 88% confidence level if cooling rate effects can be neglected.« less
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
iMAST FY2007 Annual Report

DTIC Science & Technology

2007-01-01

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

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

Spencer, Barry B.; Walker, T. B.; Bruffey, S. H.

2016-08-31

Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when themore » solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less
Capture of Tritium Released from Cladding in the Zirconium Recycle Process

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

Spencer, Barry B.; Walker, T. B.; Bruffey, Stephanie H.

2016-08-31

This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-basedmore » cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less
Nuclear reactor fuel element having improved heat transfer

DOEpatents

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

1982-03-03

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

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

Garnier, John E.; Griffith, George W.

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

Nuclear fuel element

DOEpatents

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

1980-04-29

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

NASA Astrophysics Data System (ADS)

Ott, L. J.; Robb, K. R.; Wang, D.

2014-05-01

Following the severe accidents at the Japanese Fukushima Daiichi Nuclear Power Station in 2011, the US Department of Energy initiated research and development on the enhancement of the accident tolerance of light water reactors by the development of fuels/cladding that, in comparison with the standard UO2/Zircaloy (Zr) system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. Analyses are presented that illustrate the impact of these new candidate fuel/cladding materials on the fuel performance at normal operating conditions and on the reactor system under DB and BDB accident conditions.
Microstructure evolution of recrystallized Zircaloy-4 under charged particles irradiation

NASA Astrophysics Data System (ADS)

Gaumé, M.; Onimus, F.; Dupuy, L.; Tissot, O.; Bachelet, C.; Mompiou, F.

2017-11-01

Recrystallized zirconium alloys are used as nuclear fuel cladding tubes of Pressurized Water Reactors. During operation, these alloys are submitted to fast neutron irradiation which leads to their in-reactor deformation and to a change of their mechanical properties. These phenomena are directly related to the microstructure evolution under irradiation and especially to the formation of -type dislocation loops. In the present work, the radiation damage evolution in recrystallized Zircaloy-4 has been studied using charged particles irradiation. The loop nucleation and growth kinetics, and also the helical climb of linear dislocations, were observed in-situ using a High Voltage Electron Microscope (HVEM) under 1 MeV electron irradiation at 673 and 723 K. In addition, 600 keV Zr+ ion irradiations were conducted at the same temperature. Transmission Electron Microscopy (TEM) characterizations have been performed after both types of irradiations, and show dislocation loops with a Burgers vector belonging to planes close to { 10 1 bar 0 } first order prismatic planes. The nature of the loops has been characterized. Only interstitial dislocation loops have been observed after ion irradiation at 723 K. However, after electron irradiation conducted at 673 and 723 K, both interstitial and vacancy loops were observed, the proportion of interstitial loops increasing as the temperature is increased. The loop growth kinetics analysis shows that as the temperature increases, the loop number density decreases and the loop growth rate tends to increase. An increase of the flux leads to an increase of the loop number density and a decrease of the loop growth rate. The results are compared to previous works and discussed in the light of point defects diffusion.

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

NASA Astrophysics Data System (ADS)

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

2013-11-01

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

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

Carlson, P.A.

1963-08-06

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

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

Heuser, Brent; Stubbins, James; Kozlowski, Tomasz

The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys.more » The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be. International fabrication options were explored in Europe and Asia, but this proved to be impractical, if not impossible. Consequently, experimental investigation of the Zr-Be binary system was dropped and exploration binary Zr-Y binary system was initiated. The motivation behind the Zr-Y system is the known thermodynamic stability of yttria over zirconia.« less
Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

NASA Astrophysics Data System (ADS)

Tunes, M. A.; Harrison, R. W.; Greaves, G.; Hinks, J. A.; Donnelly, S. E.

2017-09-01

Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7 ×1017ions ·cm-2 in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7 ×1017ions ·cm-2 in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.
High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

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

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

2013-01-01

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

Wang, Hong; Wang, Jy-An John

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

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

Linton, Kory D.; Burns, Zachary M.; Terrani, Kurt A.

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examinemore » postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.« 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
Fluorescent discharge lamp

NASA Technical Reports Server (NTRS)

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

1982-01-01

A rapidly illuminating fluorescent lamp 1,200 mm long and 32.5 mm in diameter with an interior conducting strip which is compatible with conventional fixtures and ballasts is described. The fluorescent lamp is composed of a linear glass tube, electrodes sealed at both ends, mercury and raregas sealed in the glass tube, a fluorescent substance clad on the inner walls of the glass tube, and a clad conducting strip extending the entire length of the glass tube in the axial direction on the inner surface of the tube.
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.
Physical and Mechanical Metallurgy of Zirconium Alloys for Nuclear Applications: A Multi-Scale Computational Study

NASA Astrophysics Data System (ADS)

Glazoff, Michael Vasily

In the post-Fukushima world, thermal and structural stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Because the nuclear industry will continue using zirconium (Zr) cladding for the foreseeable future, it becomes critical to gain a fundamental understanding of several interconnected problems. First, what are the thermodynamic and kinetic factors affecting oxidation and hydrogen pick-up by these materials at normal, off-normal conditions, and in long-term storage? Secondly, what protective coatings could be used in order to gain valuable time at off-normal conditions (temperature exceeds ~1200°C (2200°F)? Thirdly, the kinetics of the coating's oxidation must be understood. Lastly, one needs automated inspection algorithms allowing identifying cladding's defects. This work attempts to explore the problem from a computational perspective, utilizing first principles atomistic simulations, computational thermodynamics, plasticity theory, and morphological algorithms of image processing for defect identification. It consists of the four parts dealing with these four problem areas preceded by the introduction. In the 1st part, computational thermodynamics and ab initio calculations were used to shed light upon the different stages of zircaloy oxidation and hydrogen pickup, and microstructure optimization to increase thermal stability. The 2 nd part describes the kinetic theory of oxidation of the several materials considered to be perspective coatings for Zr alloys: SiC and ZrSiO4. The 3rd part deals with understanding the respective roles of the two different plasticity mechanisms in Zr nuclear alloys: twinning (at low T) and crystallographic slip (higher T's). For that goal, an advanced plasticity model was proposed. In the 4th part projectional algorithms for defect identification in zircaloy coatings are described. Conclusions and recommendations are presented in the 5th part. This integrative approach's value is in developing multi-faceted understanding of complex processes taking place in nuclear fuel rods. It helped identify several problems pertaining to the safe operations with nuclear fuel: limits of temperature that should be strictly obeyed in storage to retard zircaloy hydriding; understanding the benefits and limitations of coatings; developing in-depth understanding of Zr plasticity; developing original algorithms for defect identification in SiC-braided zircaloy. The obtained results will be useful for the nuclear industry.
Development of a (147)Pm source for beta-backscatter thickness gauge applications.

PubMed

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

2011-03-01

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

NASA Astrophysics Data System (ADS)

Shulga, A. V.

2013-03-01

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

DOE PAGES

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

2018-01-31

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

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

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

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

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

Williford, R.E.

1986-04-01

Current Emergency Core Cooling System (ECCS) Acceptance Criteria for light-water reactors include certain requirements pertaining to calculations of core performance during a Loss of Coolant Accident (LOCA). The Baker-Just correlation must be used to calculate Zircaloy-steam oxidation, calculated peak cladding temperatures (PCT) must not exceed 1204/sup 0/C, and calculated oxidation must not exceed 17% equivalent cladding reacted (17% ECR). The minimum margin of safety was estimated for each of these criteria, based on research performed in the last decade. Margins were defined as the amounts of conservatism over and above the expected extreme values computed from the data base atmore » specified confidence levels. The currently required Baker-Just oxidation correlation provides margins only over the 1100/sup 0/C to 1500/sup 0/C temperature range at the 95% confidence level. The PCT margins for thermal shock and handling failures are adequate at oxidation temperatures above 1204/sup 0/C for 210 and 160 seconds, respectively, at the 95% confidence level. ECR thermal shock and handling margins at the 50% and 95% confidence levels, respectively, range between 2% and 7% ECR for the Baker-Just correlation, but vanish at temperatures between 1100/sup 0/C and 1160/sup 0/C for the best-estimate Cathcart-Pawel correlation. Use of the Cathcart-Pawel correlation for LOCA calculations can be justified at the 85% to 88% confidence level if cooling rate effects can be neglected. 75 refs., 21 figs.« less
Electroslag Strip Cladding of Steam Generators With Alloy 690

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

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

2006-07-01

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

Fuel pin cladding

DOEpatents

Vaidyanathan, Swaminathan; Adamson, Martyn G.

1986-01-01

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

NASA Astrophysics Data System (ADS)

Arwani, M.; Kuswanto, H.

2018-04-01

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

DOEpatents

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

1959-01-27

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

DOEpatents

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

1984-01-01

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

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1986-01-28

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

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1983-12-16

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

NASA Astrophysics Data System (ADS)

Yeom, Hwasung

Experimental results investigating the feasibility of zirconium-silicide coating for accident tolerance of LWR fuel cladding coating was presented. The oxidation resistance of ZrSi2 appeared to be superior to bare Zircaloy-4 in high temperature air. It was shown that micro- and nanostructures consisting of alternating SiO2 and ZrO2 evolved during transient oxidation of ZrSi2, which was explained by spinodal phase decomposition of Zr-Si-O oxide. Coating optimization regarding oxidation resistance was performed mainly using magnetron sputter deposition method. ZrSi 2 coatings ( 3.9 microm) showed improvement of almost two orders of magnitude when compared to bare Zircaloy-4 after air-oxidation at 700 °C for 20-hours. Pre-oxidation of ZrSi2 coating at 700 °C for 5 h significantly mitigated oxygen diffusion in air-oxidation tests at 1000 °C for 1-hour and 1200 °C for 10-minutes. The ZrSi2 coating with the pre-oxidation was found to be the best condition to prevent oxide formation in Zircaloy-4 substrate in the steam condition even if the top surface of the coating was degraded by formation of zirconium-rich oxide layer. Only the ZrSiO4 phase, formed by exposing the ZrSi2 coating at 1400 °C in air, allowed for immobilization of silicon species in the oxide scale in the aqueous environments. A quench test facility was designed and built to study transient boiling heat transfer of modified Zircaloy-4 surfaces (e.g., roughened surfaces, oxidized surfaces, ZrSi2 coated surfaces) at various system conditions (e.g., elevated pressures and water subcooling). The minimum film boiling temperature increased with increasing system pressure and water subcooling, consistent with past literature. Quenching behavior was affected by the types of surface modification regardless of the environmental conditions. Quenching heat transfer was improved by the ZrSi 2 coating, a degree of surface oxidation (deltaox = 3 to 50 microm), and surface roughening (Ra 20 microm). A plausible hypothesis based on transient heat conduction models for liquid-solid contact in quenching process was proposed to explain the enhanced quenching performance. The theoretical model incorporated localized temperature behavior on superheated surface and elucidated bubble dynamics qualitatively, and predicts minimum film boiling temperature of oxidized Zirc-4 surfaces, which were in good agreement with experimental data.
COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

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

DOEpatents

Christiansen, David W.; Brown, William F.

1985-01-01

A welder for automated closure of fuel pins by a pulsed magnetic process in which the open end of a length of cladding is positioned within a complementary tube surrounded by a pulsed magnetic welder. Seals are provided at each end of the tube, which can be evacuated or can receive tag gas for direct introduction to the cladding interior. Loading of magnetic rings and end caps is accomplished automatically in conjunction with the welding steps carried out within the tube.
Study on the hydrogenation of Zircaloy-4

NASA Astrophysics Data System (ADS)

da Silva Dupim, Ivaldete; Moreira, João M. L.; Silva, Selma Luiza; Silva, Cecilia Chaves Guedes e.; Nunes, Oswaldo; Gomide, Ricardo Gonçalves

2012-08-01

In this article we investigate producing Zirconium powder from discarded Zircaloy-4 material through the hydride-dehydride method. We restrict our study to the first part of the method, namely the hydrogenation process. Differential thermal analyses of the hydrogenation process of the Zircaloy-4 show that no hydrogen absorption occurs at temperatures below 573 K and hydrogen gas pressure of 25 kPa. When the system temperature is raised to around 770 K, with the same gas pressure, the protecting oxide layer of the specimens can be overcome and they are quickly hydrogenated. The bulk of the reaction occurs in about 5 min with the precipitation of Zirconium hydrides in the Zr-δ and Zr-ɛ phases. Once the temperature passes 573 K, the incubation time to initiate the reaction is short (about 5 min). Tests in a tube furnace system with larger samples, hydrogen pressure varying from 30 to 180 kPa, and temperature from 700 to 833.15 K, show that the specimens are fully hydrogenated and can be easily pulverized. The results indicate that the hydrogenation of the Zircaloy-4 chips can be successfully undertaken at temperatures around 770 K and hydrogen gas pressure as low as 30 kPa.
Remote Field Eddy Curent Signal Modeling for the Gap Measurement of Neighboring Tubes

NASA Astrophysics Data System (ADS)

Jung, H. K.; Lee, D. H.; Lee, Y. S.

2005-04-01

The fuel channels in the Canadian Deuterium Uranium (CANDU) reactor consist of the coaxial pressure tube (PT) and the calandria tube (CT). The Liquid injection nozzle (LIN) is cross aligned with the fuel channel to control the reactor by injecting poison. For a safe operation, the gap between the LIN and CT should be maintained in order to prevent a contact of the neighboring tubes. The remote field eddy current (RFEC) method was applied to measure the gap between a nonmagnetic Zircaloy-2 liquid injection nozzle (LIN) and a Zircaloy-2 calandria tube. Under the condition of inserting the RFEC probe into the coaxial tubes and of crossing a LIN above or under the CT, the modeling of a LIN signal is needed to check the possibility of a gap measurement. The Volume Integral Code S/W which covers the axi-symmetric 3D configuration has been very rarely applied to obtain a LIN signal. This problem was solved by assuming a LIN as a flaw which can be described as a complete 3D object. This simulated LIN signal was verified by performing the laboratory experiment. The gap between the LIN and CT can be correlated with the amplitude of the LIN signals in the voltage plane. Typical noises in the fuel channel were the relative constriction, the change in the pressure tube diameter (fill-factor), thickness variation, and so on. These noise signals were simulated by using the modeling and were analyzed by considering their dependency on the phase angle and amplitude of the voltage plane in order to separate the gap signal from them. It could be concluded that the voltage plane analysis of the simulated RFEC signals were effective for obtaining the gap measurement of the neighboring tube.
Evaluation of a Zirconium Recycle Scrubber System

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

Spencer, Barry B.; Bruffey, Stephanie H.

2017-04-01

A hot-cell demonstration of the zirconium recycle process is planned as part of the Materials Recovery and Waste Forms Development (MRWFD) campaign. The process treats Zircaloy® cladding recovered from used nuclear fuel with chlorine gas to recover the zirconium as volatile ZrCl4. This releases radioactive tritium trapped in the alloy, converting it to volatile tritium chloride (TCl). To meet regulatory requirements governing radioactive emissions from nuclear fuel treatment operations, the capture and retention of a portion of this TCl may be required prior to discharge of the off-gas stream to the environment. In addition to demonstrating tritium removal from amore » synthetic zirconium recycle off-gas stream, the recovery and quantification of tritium may refine estimates of the amount of tritium present in the Zircaloy cladding of used nuclear fuel. To support these objectives, a bubbler-type scrubber was fabricated to remove the TCl from the zirconium recycle off-gas stream. The scrubber was fabricated from glass and polymer components that are resistant to chlorine and hydrochloric acid solutions. Because of concerns that the scrubber efficiency is not quantitative, tests were performed using DCl as a stand-in to experimentally measure the scrubbing efficiency of this unit. Scrubbing efficiency was ~108% ± 3% with water as the scrubber solution. Variations were noted when 1 M NaOH scrub solution was used, values ranged from 64% to 130%. The reason for the variations is not known. It is recommended that the equipment be operated with water as the scrubbing solution. Scrubbing efficiency is estimated at 100%.« less
Core thermal response and hydrogen generation of the N Reactor hydrogen mitigation design basis accident

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

White, M.D.; Lombardo, N.J.; Heard, F.J.

1988-04-01

Calculations were performed to determine core heatup, core damage, and subsequent hydrogen production of a hypothetical loss-of-cooling accident at the Department of Energy's N Reactor. The thermal transient response of the reactor core was solved using the TRUMP-BD computer program. Estimates of whole-core thermal damage and hydrogen production were made by weighting the results of multiple half-length pressure tube simulations at various power levels. The Baker-Just and Wilson parabolic rate equations for the metal-water chemical reactions modeled the key phenomena of chemical energy and hydrogen evolution. Unlimited steam was assumed available for continuous oxidation of exposed Zircaloy-2 surfaces and formore » uranium metal with fuel cladding beyond the failure temperature (1038 C). Intact fuel geometry was modeled. Maximum fuel temperatures (1181 C) in the cooled central regions of the core were predicted to occur one-half hour into the accident scenario. Maximum fuel temperatures of 1447 C occurred in the core GSCS-regions at the end of the 10-h transient. After 10-h 26% of the fuel inventory was predicted to have failed. Peak hydrogen evolution equaled 42 g/s, while 10-h integrated hydrogen evolution equaled 167 kg. 12 refs., 12 figs., 2 tabs.« 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.
Research and Engineering Operation, Irradiation Processing Department monthly record report, May 1965

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

Ambrose, T.W.

1965-06-04

Process and development activities reported include: depleted uranium irradiations, thoria irradiation, and hot die sizing. Reactor engineering activities include: brittle fracture of 190-C tanks, increased graphite temperature limits for the F reactor, VSR channel caulking, K reactor downcomer flow, zircaloy hydriding, and ribbed zircaloy process tubes. Reactor physics activities include: thoria irradiations, E-D irradiations, boiling protection with the high speed scanner, and in-core flux monitoring. Radiological engineering activities include: radiation control, classification, radiation occurrences, effluent activity data, and well car shielding. Process standards are listed, along with audits, and fuel failure experience. Operational physics and process physics studies are presented.more » Lastly, testing activities are detailed.« less
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.
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
Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

2011-06-01

Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent plastic strains are reduced; and (3) the maximum first principal stresses for certain burnup at the matrix or the cladding are lower than the ones without the hardening effect, and the differences are found to increase with burnup; and the variation rules of the interfacial stresses are similar.
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
Implications of Zircaloy creep and growth to light water reactor performance

NASA Astrophysics Data System (ADS)

Franklin, David G.; Adamson, Ronald B.

1988-10-01

Deformation of zirconium alloy components in nuclear reactors has been a concern since the decision of Admiral Rickover to use them in the US Navy submarine reactors. With the exception of the first few light water reactors (LWRs) most of the core structural materials have been fabricated from either Zircaloy-2 or Zircaloy-4. Performance of these alloys has been extremely good, even though the effects of irradiation on deformation magnitudes and mechanisms were not fully appreciated until extensive service and in-reactor tests were accomplished. Since the reactor components are designed to operate at stress levels well below yield for normal conditions, the only significant deformation is time dependent. Although creep was anticipated, the enhancement by neutron irradiation and the stress-free, nearly constant-volume shape change known as irradiation growth were not known prior to materials testing in reactors under controlled conditions. Both of these phenomena have significant impact on performance and must be accounted for properly in design. Although irradiation creep and growth have resulted in only one significant performance problem (creep collapse of fuel cladding, which has been eliminated), deformation magnitudes and, particularly, differentials in strain magnitudes, are a continuing source of interest. Factors that affect dimensional stability due to both creep and growth include temperature, fluence, residual stress, texture, and microstructure. The first two are reactor variables and the others are related to component fabrication history. This paper includes a review of the applications of Zircaloy creep and growth to LWR fuel designs, a review of the impact of in-reactor creep and growth on fuel rod and fuel assembly performance, and comments on potential improvements. Since the reactor design, fuel design and the core environment in BWRs and PWRs are quite different, appropriate separation of the application of effects are made; of course, the basic phenomena are the same in both systems.
Phase 1A Final Report for the AREVA Team Enhanced Accident Tolerant Fuels Concepts

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

Morrell, Mike E.

In response to the Department of Energy (DOE) funded initiative to develop and deploy lead fuel assemblies (LFAs) of Enhanced Accident Tolerant Fuel (EATF) into a US reactor within 10 years, AREVA put together a team to develop promising technologies for improved fuel performance during off normal operations. This team consisted of the University of Florida (UF) and the University of Wisconsin (UW), Savannah River National Laboratory (SRNL), Duke Energy and Tennessee Valley Authority (TVA). This team brought broad experience and expertise to bear on EATF development. AREVA has been designing; manufacturing and testing nuclear fuel for over 50 yearsmore » and is one of the 3 large international companies supplying fuel to the nuclear industry. The university and National Laboratory team members brought expertise in nuclear fuel concepts and materials development. Duke and TVA brought practical utility operating experience. This report documents the results from the initial “discovery phase” where the team explored options for EATF concepts that provide enhanced accident tolerance for both Design Basis (DB) and Beyond Design Basis Events (BDB). The main driver for the concepts under development were that they could be implemented in a 10 year time frame and be economically viable and acceptable to the nuclear fuel marketplace. The economics of fuel design make this DOE funded project very important to the nuclear industry. Even incremental changes to an existing fuel design can cost in the range of $100M to implement through to LFAs. If this money is invested evenly over 10 years then it can take the fuel vendor several decades after the start of the project to recover their initial investment and reach a breakeven point on the initial investment. Step or radical changes to a fuel assembly design can cost upwards of $500M and will take even longer for the fuel vendor to recover their investment. With the projected lifetimes of the current generation of nuclear power plants large scale investment by the fuel vendors is difficult to justify. Specific EATF enhancements considered by the AREVA team were; Improved performance in DB and BDB conditions; Reduced release to the environment in a catastrophic accident; Improved performance during normal operating conditions; Improved performance if US reactors start to load follow; Equal or improved economics of the fuel; and Improvements to the fuel behavior to support future transportation and storage of the used nuclear fuel (UNF). In pursuit of the above enhancements, EATF technology concepts that our team considered were; Additives to the fuel pellets which included; Chromia doping to increase fission gas retention. Chromia doping has the potential to improve load following characteristics, improve performance of the fuel pellet during clad failure, and potentially lock up cesium into the fuel matrix; Silicon Carbide (SiC) Fibers to improve thermal heat transfer in normal operating conditions which also improves margin in accident conditions and the potential to lock up iodine into the fuel matrix; Nano-diamond particles to enhance thermal conductivity; Coatings on the fuel cladding; and Nine coatings on the existing Zircaloy cladding to increase coping time and reduce clad oxidation and hydrogen generation during accident conditions, as well as reduce hydrogen pickup and mitigate hydride reorientation in the cladding. To facilitate the development process AREVA adopted a formal “Gate Review Process” (GR) that was used to review results and focus resources onto promising technologies to reduce costs and identify the technologies that would potentially be carried forward to LFAs within a 10 year period. During the initial discovery phase of the project AREVA took the decision to be relatively hands off and allow our university and National Laboratory partners to be free thinking and consider options that would not be constrained by preconceived ideas from the fuel vendor. To counter this and to keep the partners focused, the GR process was utilized. During this GR process each of the team members presented their findings to a board made up of technical experts from utilities, fuel manufacturing experts, fuel technical experts, and fuel research and development (R&D) experts. During the initial 2 years of the project there were several major accomplishments. These accomplishments, along with the implications for successfully implementing EATF, are; The experimental spark plasma sintering process (SPS) process was successfully used to produce fuel pellets containing either 10% SiC whiskers or nano-diamond particles. The ability to use this process enables the thermal margin enhancements of the fuel additives to be realized. Without the SPS process, the conventional process cannot support adding pellet additives in the required quantities; Coatings of Ti2AlC were successfully applied to Zircaloy-4 cladding. Testing of Ti2AlC coatings at Loss of Cooling Accident (LOCA) conditions showed reduced cladding oxidation compared to present un-coated Zircaloy-4 cladding. This achievement allows the presently used cladding system to be retained so that the 10 year schedule can be met. Having to implement a new cladding material will extend the development schedule beyond 10 years; Several documents were produced to support future development, testing, and licensing of EATF, including a design requirements traceability matrix, a draft business plan, a draft test plan, a draft regulatory plan, and the acceptance criteria for lead fuel assembly insertion into a commercial reactor. This preparatory work lays the foundation for ensuring the future development plans address all the areas required to test, license, and manufacture the new EATF; and In addition, the high velocity oxy-fuel and electrophoretic deposition (EPD) coating application processes were dropped from further consideration due to their inability to meet manufacturing criteria. This allows the resources to be focused on the most promising EATF concepts identified. Future development opportunities that were identified during this work include; The use of SiC or diamond requires that a new pellet production technique (Spark Plasma Sintering), be developed. This entails investment in developing, proving and implementing a new commercial pellet production process. Development of the process to apply thinner coatings is required; Coatings cannot be too “thick” or they will displace a significant volume of water in the core resulting in reduced thermal hydraulic characteristics; Application of the coating at high temperature can affect the Zircaloy substrate. This will require the development and implementation of a new cladding coating manufacturing process; and Replace the Cold Spray (CS) cladding coating application with the Physical Vapor Deposition (PVD) process to eliminate duplication of work and provide greater control over coating thicknesses. This can result in a reduction in the final cycle economic penalty of coatings.« less
Tube manufacturing and characterization of oxide dispersion strengthened ferritic steels

NASA Astrophysics Data System (ADS)

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

2000-12-01

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

DTIC Science & Technology

2007-03-01

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

DOEpatents

Bassett, C.H.

1961-05-01

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

NASA Astrophysics Data System (ADS)

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

2016-10-01

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

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

Parker, Stephen S.; White, Josh; Hosemann, Peter

High-temperature isothermal steam oxidation kinetic parameters of several ferritic alloys were determined by thermogravimetric analysis. We measured the oxidation kinetic constant (k) as a function of temperature from 900°C to 1200°C. The results show a marked increase in oxidation resistance compared to reference Zircaloy-2, with kinetic constants 3–5 orders of magnitude lower across the experimental temperature range. Our results of this investigation supplement previous findings on the properties of ferritic alloys for use as candidate cladding materials and extend kinetic parameter measurements to high-temperature steam environments suitable for assessing accident tolerance for light water reactor applications.
Oxidation Kinetics of Ferritic Alloys in High-Temperature Steam Environments

DOE PAGES

Parker, Stephen S.; White, Josh; Hosemann, Peter; ...

2017-11-03

High-temperature isothermal steam oxidation kinetic parameters of several ferritic alloys were determined by thermogravimetric analysis. We measured the oxidation kinetic constant (k) as a function of temperature from 900°C to 1200°C. The results show a marked increase in oxidation resistance compared to reference Zircaloy-2, with kinetic constants 3–5 orders of magnitude lower across the experimental temperature range. Our results of this investigation supplement previous findings on the properties of ferritic alloys for use as candidate cladding materials and extend kinetic parameter measurements to high-temperature steam environments suitable for assessing accident tolerance for light water reactor applications.
Oxidation Kinetics of Ferritic Alloys in High-Temperature Steam Environments

NASA Astrophysics Data System (ADS)

Parker, Stephen S.; White, Josh; Hosemann, Peter; Nelson, Andrew

2018-02-01

High-temperature isothermal steam oxidation kinetic parameters of several ferritic alloys were determined by thermogravimetric analysis. The oxidation kinetic constant ( k) was measured as a function of temperature from 900°C to 1200°C. The results show a marked increase in oxidation resistance compared to reference Zircaloy-2, with kinetic constants 3-5 orders of magnitude lower across the experimental temperature range. The results of this investigation supplement previous findings on the properties of ferritic alloys for use as candidate cladding materials and extend kinetic parameter measurements to high-temperature steam environments suitable for assessing accident tolerance for light water reactor applications.
Recovering tubewise power from three-dimensional nodal kinetics calculation during material relocation in an HWR

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

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

1994-12-31

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

PubMed

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

2016-04-18

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

NASA Astrophysics Data System (ADS)

Gaillac, Alexis; Ly, Céline

2018-05-01

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

DOEpatents

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

1992-01-01

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

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

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

1986-01-01

This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment wasmore » heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.« less
3D-Printed Broadband Dielectric Tube Terahertz Waveguide with Anti-Reflection Structure

NASA Astrophysics Data System (ADS)

Vogt, Dominik Walter; Leonhardt, Rainer

2016-11-01

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

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

Patra, Anirban; Tome, Carlos

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

NASA Astrophysics Data System (ADS)

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

2016-09-01

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

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

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

1998-03-01

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

RIA simulation tests using driver tube for ATF cladding

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

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

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

NASA Astrophysics Data System (ADS)

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

2017-08-01

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

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

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

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300–350 °C under a representative high heat flux (~0.66 MW/m 2) during one cycle of irradiation in an un-instrumented “rabbit” capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb themore » expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. Furthermore, the success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.« less
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
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.
Corrosion evaluation of N reactor pressure tube 1756

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

Larrick, A.P.

1967-10-26

N Reactor Zircaloy-2 pressure tube No. 1756 and its associated ASTM A234 steel nozzles were examined for corrosion and hydrogen content after approximately 300 days exposure in-reactor. Visual examination showed tight, adherent, dull black oxides in the pressure tube except for scratching in the bottom due to sliding of fuel and fuel spacers through the tube during charge- discharge operations. Several fretted areas up to $sup 3$/$sub 8$ inch wide by $sup 1$/$sub 2$ inch long by up to 13 mils deep were observed at the downstream end--these pits were caused by vibration of the fuel spacers against the pressuremore » tube. Hydrogen levels were fairly constant along the tube length with an average of about 19 +- 6 ppm except at one location. At approximately 30 inches from the front end of the tube a sharp peak to a maximum of 58 ppm hydrogen occurred. The reason for the peak is unknown. (auth)« less
Understanding thermally activated plastic deformation behavior of Zircaloy-4

NASA Astrophysics Data System (ADS)

Kumar, N.; Alomari, A.; Murty, K. L.

2018-06-01

Understanding micromechanics of plastic deformation of existing materials is essential for improving their properties further and/or developing advanced materials for much more severe load bearing applications. The objective of the present work was to understand micromechanics of plastic deformation of Zircaloy-4, a zirconium-based alloy used as fuel cladding and channel (in BWRs) material in nuclear reactors. The Zircaloy-4 in recrystallized (at 973 K for 4 h) condition was subjected to uniaxial tensile testing at a constant cross-head velocity at temperatures in the range 293 K-1073 K and repeated stress relaxation tests at 293 K, 573 K, and 773 K. The minimum in the total elongation was indicative of dynamic strain aging phenomenon in this alloy in the intermediate temperature regime. The yield stress of the alloy was separated into effective and athermal components and the transition from thermally activated dislocation glide to athermal regime took place at around 673 K with the athermal stress estimated to be 115 MPa. The activation volume was found to be in the range of 40 b3 to 160 b3. The activation volume values and the data analyses using the solid-solution models in literature indicated dislocation-solute interaction to be a potential deformation mechanism in thermally activated regime. The activation energy calculated at 573 K was very close to that found for diffusivity of oxygen in α-Zr that was suggestive of dislocations-oxygen interaction during plastic deformation. This type of information may be helpful in alloy design in selecting different elements to control the deformation behavior of the material and impart desired mechanical properties in those materials for specific applications.
Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

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

Ott, Larry J.; Howell, Michael; Robb, Kevin R.

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristicsmore » are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate constant (3 times and 10 times that of the rate constant for APMT) had a negligible impact on the early stages of the accident and minor impacts on the accident progression after the first relocation of the fuel. At temperatures below 1,500°C, increasing the rate constant for APMT by a factor of 10 still resulted in only minor FeCrAl oxidation. In general, the gains afforded by the FeCrAl enhanced ATF concept with respect to accident sequence timing and combustible gas generation are consistent with previous efforts. Compared with the traditional Zircaloy-based cladding and channel box system, the FeCrAl concept could provide a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. For example, a station blackout was simulated in which cooling water injection was lost 36 hours after shutdown. The timing to first fuel relocation was delayed by approximately 5 h for the FeCrAl ATF concept compared with that of the traditional Zircaloy-based cladding and channel box system.« less
Fabrication of seamless calandria tubes by cold pilgering route using 3-pass and 2-pass schedules

NASA Astrophysics Data System (ADS)

Saibaba, N.

2008-12-01

Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90-95. Such tubes are conventionally produced by the 'welded route', which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the 'seamless route' which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties.
Tapered fiber Mach-Zehnder interferometers for vibration and elasticity sensing applications.

PubMed

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

2013-05-06

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

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

Yang, Yong; Phillpot, Simon

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

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

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

2014-08-30

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

Jensen, Colby B.; Folsom, Charles P.; Davis, Cliff B.

Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operatingmore » pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.« less
Survey of Thermal-Fluids Evaluation and Confirmatory Experimental Validation Requirements of Accident Tolerant Cladding Concepts with Focus on Boiling Heat Transfer Characteristics

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

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

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance, also known as accident tolerant fuel (ATF). Thermal-fluids characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One vital characteristic related to boiling heat transfer is the critical heat flux (CHF). CHF plays a vital role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. This deliverable is a scoping survey of thermal-fluids evaluation andmore » confirmatory experimental validation requirements of accident tolerant cladding concepts with a focus on boiling heat transfer characteristics. The key takeaway messages of this report are: 1. CHF prediction accuracy is important and the correlations may have significant uncertainty. 2. Surface conditions are important factors for CHF, primarily the wettability that is characterized by contact angle. Smaller contact angle indicates greater wettability, which increases the CHF. Surface roughness also impacts wettability. Results in the literature for pool boiling experiments indicate changes in CHF by up to 60% for several ATF cladding candidates. 3. The measured wettability of FeCrAl (i.e., contact angle and roughness) indicates that CHF should be investigated further through pool boiling and flow boiling experiments. 4. Initial measurements of static advancing contact angle and surface roughness indicate that FeCrAl is expected to have a higher CHF than Zircaloy. The measured contact angle of different FeCrAl alloy samples depends on oxide layer thickness and composition. The static advancing contact angle tends to decrease as the oxide layer thickness increases.« less
Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

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

Shaver, Mark W.; Lanning, Donald D.

2010-02-01

The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum ofmore » the individual components equaling the measured values.« less
Process development and fabrication of space station type aluminum-clad graphite epoxy struts

NASA Technical Reports Server (NTRS)

Ring, L. R.

1990-01-01

The manufacture of aluminum-clad graphite epoxy struts, designed for application to the Space Station truss structure, is described. The strut requirements are identified, and the strut material selection rationale is discussed. The manufacturing procedure is described, and shop documents describing the details are included. Dry graphite fiber, Pitch-75, is pulled between two concentric aluminum tubes. Epoxy resin is then injected and cured. After reduction of the aluminum wall thickness by chemical milling the end fittings are bonded on the tubes. A discussion of the characteristics of the manufactured struts, i.e., geometry, weight, and any anomalies of the individual struts is included.
Tensile properties and flow behavior analysis of modified 9Cr-1Mo steel clad tube material

NASA Astrophysics Data System (ADS)

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

2014-11-01

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

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

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

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

Microstructure studies of interdiffusion behavior of U 3Si 2/Zircaloy-4 at 800 and 1000 °C

DOE PAGES

He, Lingfeng; Harp, Jason M.; Hoggan, Rita E.; ...

2017-01-22

Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions when in contact with its cladding. As new fuel types are developed, it is crucial to understand the interaction behavior between fuel and its cladding. Diffusion experiments between U 3Si 2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000°C up to 100 hours. The microstructure of pristine U 3Si 2 and U 3Si 2/Zry-4 interdiffusion products were examined using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) equipped with an energy dispersive X-ray spectroscopy (EDS) system. The primary interdiffusion product observed at 800°C is ZrSi 2,more » with secondary phases of U-Zr in the Zry-4, and Fe-Cr-W-Zr-Si phases at Zry-4/ZrSi 2 interface and Fe-Cr-U-Si phases at ZrSi 2/U-Si interface. As a result, the primary interdiffusion products at 1000°C were Zr 2Si, U-Zr-Fe-Ni, U, U-Zr, and a low melting point phase U 6Fe.« less
Uniform corrosion of FeCrAl alloys in LWR coolant environments

NASA Astrophysics Data System (ADS)

Terrani, K. A.; Pint, B. A.; Kim, Y.-J.; Unocic, K. A.; Yang, Y.; Silva, C. M.; Meyer, H. M.; Rebak, R. B.

2016-10-01

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300-500 μm thick cladding.
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
Uniform corrosion of FeCrAl alloys in LWR coolant environments

DOE PAGES

Terrani, K. A.; Pint, B. A.; Kim, Y. -J.; ...

2016-06-29

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation ofmore » very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. Finally, the maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ~2 μm, which is inconsequential for a ~300–500 μm thick cladding.« 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.
Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

DOEpatents

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

1982-01-01

An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.
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
Implementation of DoD ManTech Projects Receiving FY03-05 Funds

DTIC Science & Technology

2008-12-01

ManTech project is providing an integrated tool to grind, laser clad , and finish repair work of submarine VLS tubes. The system is scheduled for...hours from about 400 - 500 hours to 40 - 50 hours resulting in improved readiness. The in- situ laser clad process eliminates the use of hazardous...Alternate Remote Shipboard Lighting for Reduced Costs .......................................................... 14 2.2.10 ManTech for Military Lasers
Studies of electrochemical oxidation of Zircaloy nuclear reactor fuel cladding using time-of-flight-energy elastic recoil detection analysis

NASA Astrophysics Data System (ADS)

Whitlow, H. J.; Zhang, Y.; Wang, Y.; Winzell, T.; Simic, N.; Ahlberg, E.; Limbäck, M.; Wikmark, G.

2000-03-01

The trend towards increased fuel burn-up and higher operating temperatures in order to achieve more economic operation of nuclear power plants places demands on a better understanding of oxidative corrosion of Zircaloy (Zry) fuel rod cladding. As part of a programme to study these processes we have applied time-of-flight-energy elastic recoil detection (ToF-E ERD), electrochemical impedance measurements and scanning electron microscopy to quantitatively characterise thin-oxide films corresponding to the pre-transition oxidation regime. Oxide films of different nominal thickness in the 9-300 nm range were grown on a series of rolled Zr and Zry-2 plates by anodisation in dilute H 2SO 4 with applied voltages. The dielectric thickness of the oxide layer was determined from the electrochemical impedance measurements and the surface topography characterised by scanning electron microscopy. ToF-E ERD with a 60 MeV 127I 11+ ion beam was used to determine the oxygen content and chemical composition of the oxide layer. In the Zr samples, the oxygen content (O atom cm -2) that was determined by ERD was closely similar to the O content derived from impedance measurements from the dielectric film. The absolute agreement was well within the uncertainty associated with the stopping powers. Moreover, the measured composition of the thick oxide layers corresponded to ZrO 2 for the films thicker than 65 nm where the oxide layer was resolved in the ERD depth profile. Zry-2 samples exhibited a similar behaviour for small thickness ( ⩽130 nm) but had an enhanced O content at larger thicknesses that could be associated either with enhanced rough surface topography or porous oxide formation that was correlated with the presence of Second Phase Particles (SPP) in Zry-2. The concentration of SPP elements (Fe, Cr, Ni) in relation to Zr was the same in the outer 9×10 17 atom cm -2 of oxide as in the same thickness of metal. The results also revealed the presence of about 1 at.% 32S in the oxides on the Zr and Zry-2 samples which presumably originates from the electrolyte.
Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

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

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

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

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

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

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

2016-10-25

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

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

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

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

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

Jason Hales; Various

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

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

Robb, Kevin R

2015-01-01

Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditionalmore » Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.« 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
Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

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

Gamble, Kyle A.; Hales, Jason D.

2016-12-01

The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of themore » concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. Sensitivity analyses are completed using Sandia National Laboratories’ DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.« less
Modelling Accident Tolerant Fuel Concepts

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

Hales, Jason Dean; Gamble, Kyle Allan Lawrence

2016-05-01

The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether eithermore » of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced Fuels Campaign.« less
Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

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

Patra, Anirban; Tomé, Carlos N.

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

DOE PAGES

Patra, Anirban; Tomé, Carlos N.

2017-03-06

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

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

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

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

1998-07-01

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

DOEpatents

Glatter, J.; Schaner, B.E.

1961-07-14

A process of forming high-density compacts of a powdered ceramic material is described by agglomerating the powdered ceramic material with a heat- decompossble binder, adding a heat-decompossble lubricant to the agglomerated material, placing a quantity of the material into a die cavity, pressing the material to form a compact, pretreating the compacts in a nonoxidizing atmosphere to remove the binder and lubricant, and sintering the compacts. When this process is used for making nuclear reactor fuel elements, the ceramic material is an oxide powder of a fissionsble material and after forming, the compacts are placed in a cladding tube which is closed at its ends by vapor tight end caps, so that the sintered compacts are held in close contact with each other and with the interior wall of the cladding tube.
Estimation of carbon 14 inventory in hull and end-piece wastes from Japanese commercial reprocessing operation

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

Tomofumi Sakuragi; Hiromi Tanabe; Emiko Hirose

2013-07-01

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the {sup 14}N(n,p){sup 14}C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated bymore » ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types x 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46x10{sup 14} Bq, consisting of 2.58x10{sup 14} Bq for BWRs and 1.87x10{sup 14} Bq for PWRs, and is consistent with the safety assessment of 4.4x10{sup 14} Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72x10{sup 13} Bq (13% of the total) in the previous assessment to 1.30x10{sup 13} Bq (2.9% of the total; consisting of 1.48x10{sup 12} for BWRs and 1.15x10{sup 13} for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible. (authors)« less
METHOD FOR MAKING FUEL ELEMENTS

DOEpatents

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

1960-08-01

A method is given for making zirconium-clad uranium wire. A tube of zirconium is closed with a zirconium plug, after which a chilled uranium core is inserted in the tube to rest against the plug. Additional plugs and cores are inserted alternately as desired. The assembly is then sheathed with iron, hot worked to the desired size, and the iron sheath removed.
Tensile properties from room temperature to 1315 C of tungsten-lined tantalum-alloy (T-111) tubing fabricated by hot isostatic pressing

NASA Technical Reports Server (NTRS)

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

1974-01-01

The effects were studied of a thin tungsten liner on the tensile properties of T-111 tubing considered for fuel cladding in a space power nuclear reactor concept. The results indicate that the metallurgically bonded liner had no appreciable effects on the properties of the T-111 tubing. A hot isostatic pressing method used to apply the liners is described along with a means for overcoming the possible embrittling effects of hydrogen contamination.
History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

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

Larry Zirker; Nathan Jerred; Dr. Indrajit Charit

2012-03-01

Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.
Data summary report for fission product release test VI-6

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

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

Test VI-6 was the sixth test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium. The fuel had experienced a burnup of {approximately}42 MWd/kg, with inert gas release during irradiation of {approximately}2%. The fuel specimen was heated in an induction furnace at 2300 K for 60 min, initially in hydrogen, then in a steam atmosphere. The released fission products were collected in three sequentially operated collection trains designed to facilitate sampling and analysis. The fission product inventories in the fuel were measured directlymore » by gamma-ray spectrometry, where possible, and were calculated by ORIGEN2. Integral releases were 75% for {sup 85}Kr, 67% for {sup 129}I, 64% for {sup 125}Sb, 80% for both {sup 134}Cs and {sup 137}Cs, 14% for {sup 154}Eu, 63% for Te, 32% for Ba, 13% for Mo, and 5.8% for Sr. Of the totals released from the fuel, 43% of the Cs, 32% of the Sb, and 98% of the Eu were deposited in the outlet end of the furnace. During the heatup in hydrogen, the Zircaloy cladding melted, ran down, and reacted with some of the UO{sub 2} and fission products, especially Te and Sb. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.57 g, almost equally divided between thermal gradient tubes and filters. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL Diffusion Model.« less
Layer Protecting the Surface of Zirconium Used in Nuclear Reactors.

PubMed

Ashcheulov, Petr; Skoda, Radek; Skarohlíd, Jan; Taylor, Andrew; Fendrych, Frantisek; Kratochvílová, Irena

2016-01-01

Zirconium alloys have very useful properties for nuclear facilities applications having low absorption cross-section of thermal electrons, high ductility, hardness and corrosion resistance. However, there is also a significant disadvantage: it reacts with water steam and during this (oxidative) reaction it releases hydrogen gas, which partly diffuses into the alloy forming zirconium hydrides. A new strategy for surface protection of zirconium alloys against undesirable oxidation in nuclear reactors by polycrystalline diamond film has been patented- Czech patent 305059: Layer protecting the surface of zirconium alloys used in nuclear reactors and PCT patent: Layer for protecting surface of zirconium alloys (Patent Number: WO2015039636-A1). The zirconium alloy surface was covered by polycrystalline diamond layer grown in plasma enhanced chemical vapor deposition apparatus with linear antenna delivery system. Substantial progress in the description and understanding of the polycrystalline diamond/ zirconium alloys interface and material properties under standard and nuclear reactors conditions (irradiation, hot steam oxidation experiments and heating-quenching cycles) was made. In addition, process technology for the deposition of protective polycrystalline diamond films onto the surface of zirconium alloys was optimized. Zircaloy2 nuclear fuel pins were covered by 300 nm thick protective polycrystalline diamond layer (PCD) using plasma enhanced chemical vapor deposition apparatus with linear antenna delivery system. The polycrystalline diamond layer protects the zirconium alloy surface against undesirable oxidation and consolidates its chemical stability while preserving its functionality. PCD covered Zircaloy2 and standard Zircaloy2 pins were for 30 min. oxidized in 1100°C hot steam. Under these conditions α phase of zirconium changes to β phase (more opened for oxygen/hydrogen diffusion). PCD anticorrosion protection of Zircaloy nuclear fuel assemblies can significantly prolong lifetime of Zirconium alloy in nuclear reactors even above Zirconium phase transition temperatures. Even after ion beam irradiation (10 dpa, 3 MeV Fe(2+)) the diamond film still shows satisfactory structural integrity with both sp(3) and sp(2) carbon phases. Zircaloy2 under the carbon-based protective layer after hot steam oxidation test differed from the original Zircaloy2 material composition only very slightly, proving that the diamond coating increases the material resistance to high temperature oxidation. Zirconium alloys nuclear fuel pins' surfaces were covered by compact and homogeneous polycrystalline diamond layers consisting of sp(3) and sp(2) carbon phases with a high crystalline diamond content and low roughness. Diamond withstands very high temperatures, has excellent thermal conductivity and low chemical reactivity, it does not degrade over time and (important for the nuclear fuel cladding) being pure carbon, it has perfect neutron cross-section properties. Moreover, polycrystalline diamond layers consisting of crystalline (sp(3)) and amorphous (sp(2)) carbon phases could have suitable thermal expansion. Zirconium alloys coated with polycrystalline diamond film are protected against undesirable changes and processes. Further, the polycrystalline diamond layer prevents the reaction between the alloy surface and water vapor. During such reaction, water molecules dissociate and initiate formation of zirconium dioxide and hydrogen, accompanied by the release of large amount of heat. Thus the protective layer prevents the formation of hydrogen and the release of reaction heat. Few relevant patents to the topic have been reviewed and cited.
Fully Ceramic Microencapsulated Fuel Development for LWR Applications

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

Snead, Lance Lewis; Besmann, Theodore M; Terrani, Kurt A

2012-01-01

The concept, fabrication, and key feasibility issues of a new fuel form based on the microencapsulated (TRISO-type) fuel which has been specifically engineered for LWR application and compacted within a SiC matrix will be presented. This fuel, the so-called fully ceramic microencapsulated fuel is currently undergoing development as an accident tolerant fuel for potential UO2 replacement in commercial LWRs. While the ability of this fuel to facilitate normal LWR cycle performance is an ongoing effort within the program, this will not be a focus of this paper. Rather, key feasibility and performance aspects of the fuel will be presented includingmore » the ability to fabricate a LWR-specific TRISO, the need for and route to a high thermal conductivity and fully dense matrix that contains neutron poisons, and the performance of that matrix under irradiation and the interaction of the fuel with commercial zircaloy clad.« less
Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

NASA Astrophysics Data System (ADS)

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

2018-05-01

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

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

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

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

DOE PAGES

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

2018-03-02

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

DOEpatents

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

2004-10-19

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

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

Wang, Jy-An John

To determine the hoop tensile properties of irradiated fuel cladding in a hot cell, a cone wedge ring expansion test method was developed. A four-piece wedge insert was designed with tapered angles matched to the cone shape of a loading piston. The ring specimen was expanded in the radial direction by the lateral expansion of the wedges under the downward movement of the piston. The advantages of the proposed method are that implementation of the test setup in a hot cell is simple and easy, and that it enables a direct strain measurement of the test specimen from the piston’smore » vertical displacement soon after the wedge-clad contact resistance is initiated.« less
Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

NASA Astrophysics Data System (ADS)

Carroll, Spencer

As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel performance code developed by PNNL and used by the Nuclear Regulatory Commission (NRC) as a licensing code for US reactors. FRAPCON will give insight into how these new fuel-cladding combinations will affect cladding hoop stress and help determine if the new materials are feasible for use in a reactor. To accurately simulate the interaction between the new materials, a soft pellet model that allows for stresses on the pellet to affect pellet deformation will have to be implemented. Currently, FRAPCON uses a rigid pellet model that does not allow for feedback of the cladding onto the pellet. Since SiC does not creep at the temperatures being considered and is not ductile, any PCMI create a much higher interfacial pressure than is possible with Zircaloy. Because of this, it is necessary to implement a model that allows for pellet creep to alleviate some of these cladding stresses. These results will then be compared to FEMAXI-6, a Japanese fuel performance code that already calculates pellet stress and allows for cladding feedback onto the pellet. This research is intended to be a continuation and verification of previous work done by USC on the analysis of accident tolerant fuels with alternative claddings and is intended to prove that a soft pellet model is necessary to accurately model any fuel with SiC cladding.
DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

DOE PAGES

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

2015-02-27

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

NASA Astrophysics Data System (ADS)

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

2015-02-01

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

Uranium nitride fuel fabrication for SP-100 reactors

NASA Technical Reports Server (NTRS)

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

1987-01-01

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.
Uranium nitride fuel fabrication for SP-100 reactors

NASA Astrophysics Data System (ADS)

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

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

NASA Astrophysics Data System (ADS)

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

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.
(Project 14-6770) An Investigation to Establish Multiphysical Property Dataset of Nuclear Materials Based on in-situ Observations and Measurements

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

Tomar, Vikas; Haque, Aman; Hattar, Khalid

In-core nuclear materials including fuel pins and cladding materials fail due to issues including corrosion, mechanical wear, and pellet cladding interaction. In most such scenario microstructure dependent and corrosioninduced chemistry dependent property changes significantly affect performance of cladding, pellet, and housing. Emphasis of this work was on replace conventional pellet-cladding material models with a new straingradient viscoplasticity model that is informed by transmission electron microscopy (TEM) based measurements and by nanomechanical Raman spectroscopy (NMRS) based measurements. The TEM measurements are quantitative in nature and therefore reveal stress-strain relations with simultaneous insights into mechanisms of deformation at nanoscale. The NMRS measurementsmore » reveal the similar information at mesoscale along with additional information on relating local microstructural stresses with applied stresses. The resulting information is used to fit constants in the strain gradient viscoplasticity model as well as to validate one. During TEM measurements, a micro-electro-mechanical system based setup was developed with mechanical actuation, sensing, heating, and electrical loading. Contrary to post-mortem analysis or qualitative visualization, this setup combines direct visualization of the mechanisms behind deformation with measurement of stress, strain, thermal and electrical properties. The unique research philosophy of visualizing the microstructure at high resolution while measuring the properties led to fundamental understanding in grain size and temperature effects on measured mechanical properties such as fracture toughness. A key contribution is the role of mechanical loading boundary conditions to deconvolute the insitu TEM based nanoscale and NMRS based mesoscale data to bulk behavior. First the literature based pellet cladding mechanical interaction model based on the work of Retel’s and Williamson’s in literature work to predict tempurature and stress distribution in cladding and pellet at normal operating condition was analyzed. Later the data was fitted to find constants for a viscoplastic strain gradient model. The developed model still needs to be refined and calibrated using various experimental results. That remains the focus of future work. Overall, a major thrust of the work was therefore on active control of the microstructure (grain size, defect density and types) exploiting the multi-physics coupling in materials. In particular, using experiments the synergy of current density, mechanical stress and temperature were studied to annihilate defects and recrystallize grains. The developed model is being examined for implementation in BISON. Multiple invited talks, international journal publications, and conference publications were performed by students supported on this work. Another output is support multiple PhD and masters thesis students who will be an important asset for future basic nuclear research. Future Work Recommendations: A nuclear reactor operates under significant variations of thermal loads due to energy cycling and mechanical loads due to constraint effects. Significant thermal and chemical diffusion takes place at the pallet-cladding level. While the proposed work established new experimental approach and new dataset for Zircaloy-4, the irradiation level was in the range of 1-2 dpa. Samples with higher dpa need to be examined. Therefore, a continual of support of the performed work is essential. Currently, these are the only experiments that can measure the produced data. The work also needs to be extended to different fuel types and cladding types such as SiC and FeCrAl based claddings. A combination of datasets for these materials can then be used to analyze accurately predict behavior of critical pellet cladding systems in accident scenario with high heat flux and high thermal loads. This is a BIG unknown as if now.« less
High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

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

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

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

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

Ang, Caen K.; Burns, Joseph R.; Terrani, Kurt A.

2016-09-01

There is a need to increase the safety margins of current and future light water reactors (LWRs) due to the unfortunate events at Fukushima Daiichi Nuclear Plant. Safety is crucial to restore public confidence in nuclear energy, acknowledged as an economical, high-density energy solution to climate change. The development of accident-tolerant fuel (ATF) concepts is crucial to this endeavor. The objective of ATF is to delay the consequences of accident progression, being inset in high temperature steam and maintaining high thermomechanical strength for radionuclide retention. The use of advanced SiCf-SiC composite as a substitute for zircaloy-based cladding is being considered.more » However, at normal operations, SiC is vulnerable to the reactor coolant and may corrode at an unacceptable rate. As a ceramic-matrix composite material, it is likely to undergo microcracking operation, which may compromise the ability to contain gaseous fission products. A proposed solution to both issues is the application of mitigation coatings for use in normal operations. At Oak Ridge National Laboratory (ORNL), three coating technologies have been investigated with industry collaborators and vendors. These are electrochemical deposition, cathodic arc physical vapor deposition (PVD hereafter) and vacuum plasma spray (VPS). The objective of this document is to summarize these processing technologies, the resultant as-processed microstructures and properties of the coatings. In all processes, substrate constraint resulted in substantial tensile stresses within the coating layer. Each technology must mitigate this tensile stress. Electrochemical coatings use chromium as the coolant facing material, and are deposited on a nickel or carbon “bond coat”. This is economical but suffers microcracking in the chromium layer. PVD-based coatings use chromium and titanium in both metallic form and nitrides, and can be deposited defense-in-depth as multilayers. This vapor method eliminated tensile stress during processing and coatings were up to ~30 μm thick without microcracking. VPS produced coatings based on Zircaloy-2, which has a proven reactor-compatibility. The tensile stresses appearred to be partially mitigated by annealing. Analysis showed that VPS coatings required further optimizations to prevent adverse reactions with the substrate and need a minimum thickness of ~50 μm. In addition, development of coatings are constrained by neutronic depletion analysis, which clearly indicated enrichment as an issue if the coating is too thick. Based on the present work, the cathodic arc PVD technology was considered ready for the extensive testing and evaluation on cladding materials due to their ability to mitigate the excessive tensile stresses and the reasonable coating quality achieved. The VPS Zircaloy-2 coating technology required additional development toward mitigation of issues related to the substrate reaction and porosity. In the future, PVD and VPS will have be scaled upon successful development and demonstration. Electrochemical coatings, which are proven scalability, currently require development to mitigate issues related to the tensile stress after deposition.« 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.
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
Stressed waveguides with tubular depressed-cladding inscribed in phosphate glasses by femtosecond hollow laser beams.

PubMed

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

2012-08-01

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

Oxide Dispersion Strengthened (ODS) steels are foreseen as fuel cladding material in the coming generation of Sodium Fast Reactors (SFR). Cladding tubes are manufactured by hot extrusion and subsequent cold forming steps. In this study, a 9 wt% Cr ODS steel exhibiting α-γ phase transformation at high temperature is cold formed under industrial conditions with a large section reduction in two pilgering steps. The influence of pilgering process parameters and intermediate heat treatment on the microstructure evolution is studied experimentally using Electron Backscattering Diffraction (EBSD) and X-ray Diffraction (XRD) methods. Pilgered samples show elongated grains and a high texture formation with a preferential orientation along the rolling direction. During the heat treatment, grain morphology is recovered from elongated grains to almost equiaxed ones, while the well-known α-fiber texture presents an unexpected increase in intensity. The remarkable temperature stability of this fiber is attributed to a crystallographic structure memory effect during phase transformations.
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
CVTR PROJECT. CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC. MONTHLY PROGRESS REPORT, MAY 1961

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

None

1961-10-31

The capsule A-2 was removed from the WTR reflector hole at the end of the WTR Cycle 13, and was stored in the WTR canal. The in-pile loop has operated for eight months and the test thimble was irradiated a total of 108 days. Tensile tests were completed on the extruded and annealed Zircaloy-4 Phase-II pressure tubes. The tensile properties varied with location in the pressure tube. The lowest values were obtained in the top flange where the material was fully annealed for ten hours at 800 deg C. Increased properties were achieved from working the material during extrusion operations.more » A shielding ring is provided to prevent streaming through a void generated by the rotating shield volley supports. It was determined that an additional thickness of iron or steel is required to compensate for the loss of shielding from the removal of one foot of concrete at the bottom of the trench. Various portions of the U-tube and fuel assemblies were homogenized in various axial regions for computer studies. The studies indicated a decrease of 500 hours in core life from non-uniform axial burnup. Pressure tube specimens are being tested under the impulsive test burst program. A test specimen experienced a 51% increase in O.D. under 20 impact blows before it failed. Observations of the tested specimens indicated ductilities far in excess of those predicted from the material's behavior in uniaxial tension. Teste on a Zircaloy-stainless steel joint were concluded after an extensive program of testing under various pressure, temperature and bending moment conditions. No sign of leakage was noted throughout the program. Subsequent inspection of the joint showed cracks in the sleeve portion of the joint. Analysis of the test water indicated a chloride content of approx 88 ppm. A test fuel assembly was dismantled and converted to a four baffle design. Modifications were made to the prototype control-rod-drive system. The alignment between ths vertical and horizontal miter gears was improved by charging the mounting of horizontal shaft and bearings. Scram tests were resumed; these tests indicated that the dashpot was acting too soon. The dashpot is being modified. (auth)« less
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)
Wide angle near-field optical probes by reverse tube etching.

PubMed

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

2006-04-01

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

NASA Astrophysics Data System (ADS)

Hermann, David J.; Gorenflo, Ronald L.

1998-09-01

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

DOEpatents

Korton, George

2004-02-24

This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.
Sliding wear and friction behaviour of zircaloy-4 in water

NASA Astrophysics Data System (ADS)

Sharma, Garima; Limaye, P. K.; Jadhav, D. T.

2009-11-01

In water cooled nuclear reactors, the sliding of fuel bundles in fuel channel handling system can lead to severe wear and it is an important topic to study. In the present study, sliding wear behaviour of zircaloy-4 was investigated in water (pH ˜ 10.5) using ball-on-plate sliding wear tester. Sliding wear resistance zircaloy-4 against SS 316 was examined at room temperature. Sliding wear tests were carried out at different load and sliding frequencies. The coefficient of friction of zircaloy-4 was also measured during each tests and it was found to decrease slightly with the increase in applied load. The micro-mechanisms responsible for wear in zircaloy-4 were identified to be microcutting, micropitting and microcracking of deformed subsurface zones in water.
Viability of thin wall tube forming of ATF FeCrAl

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

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

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

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

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

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

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

Kalimullah

1994-03-01

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

NASA Astrophysics Data System (ADS)

Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

2016-12-01

This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.
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
Design and analysis of the radiator structure for space power systems

NASA Technical Reports Server (NTRS)

Dauterman, W. H.; Montgomery, L. D.

1973-01-01

The design, analysis, fabrication, and development of the 5-kWe radiator structure are shown. Thermal performance, meteoroid protection, structural capability during launch, development testing and space operation, material evaluation, and the configuration selection are described. The fin-tube development program depends on the relative values of the thermal coefficients of expansion. The initial selection of aluminum fins and Type 316 stainless-steel tubes was based on previous experience; however, the large differential in their expansion rates showed that an alternate, more compatible, combination was needed. Copper, stainless-steel-clad copper, boron-impregnated aluminum, and an independent radiator with a titanium structure were all considered as alternate materials. The final selection was Lockalloy fins with Type 304 stainless-steel D tubes.
CUTTING AND WEDGING JACKET REMOVER

DOEpatents

Freedman, M.; Raynor, S.

1959-04-01

A tool is presented for stripping cladded jackets from fissionable fuel elements. The tool is a tube which fits closely around the jacket and which has two cutting edges at opposite sides of one end. These cutting edges are adjusted to penetrate only the jacket so that by moving the edges downward the jacket is cut into two pieces.
Modeling Explosive Cladding of Metallic Liners to Gun Tubes

DTIC Science & Technology

2010-01-01

a Jones Wilkins Lee ( JWL ) equation of state was parameterized using nonlinear optimization (ref. 8) and scaling the empirical v2£ for other volume...expansions based on TNT. The JWL equation of state is ( ,. A i -RiV* , GJE RiV* V IS./K v f +7~* (2) where P is pressure, V is
Mechanical and thermal properties of bulk ZrB2

NASA Astrophysics Data System (ADS)

Nakamori, Fumihiro; Ohishi, Yuji; Muta, Hiroaki; Kurosaki, Ken; Fukumoto, Ken-ichi; Yamanaka, Shinsuke

2015-12-01

ZrB2 appears to have formed in the fuel debris at the Fukushima Daiichi nuclear disaster site, through the reaction between Zircaloy cladding materials and the control rod material B4C. Since ZrB2 has a high melting point of 3518 K, the ceramic has been widely studied as a heat-resistant material. Although various studies on the thermochemical and thermophysical properties have been performed for ZrB2, significant differences exist in the data, possibly due to impurities or the porosity within the studied samples. In the present study, we have prepared a ZrB2 bulk sample with 93.1% theoretical density by sintering ZrB2 powder. On this sample, we have comprehensively examined the thermal and mechanical properties of ZrB2 by the measurement of specific heat, ultrasonic sound velocities, thermal diffusivity, and thermal expansion. Vickers hardness and fracture toughness were also measured and found to be 13-23 GPa and 1.8-2.8 MPa m0.5, respectively. The relationships between these properties were carefully examined in the present study.
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
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.
Magnetic field sensing based on tilted fiber Bragg grating coated with nanoparticle magnetic fluid

NASA Astrophysics Data System (ADS)

Yang, Dexing; Du, Lei; Xu, Zengqi; Jiang, Yajun; Xu, Jian; Wang, Meirong; Bai, Yang; Wang, Haiyan

2014-02-01

A magnetic field sensor based on a tilted fiber Bragg grating (TFBG) coated with magnetic fluid is proposed and demonstrated experimentally. The sensing element is made by injecting the magnetic fluid into a capillary tube which contains a TFBG. The resonant wavelengths of the cladding modes of TFBG shift by varying the magnetic field which is perpendicular to the axis of TFBG. The results indicate that the resonant wavelength shifts of the cladding modes show a nonlinear dependence on the magnetic field. As the magnetic field increases to 32 mT, the largest resonant wavelength shift reaches to 106 pm. Moreover, this sensor shows good repeatability when it is used for magnetic field sensing.
Aspects of forming metal-clad melt-processed Y-Ba-Cu-O tapes

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

Kozlowski, G.; Oberly, C.E.; Ho, J.

1991-03-01

This paper reports on melt-processing of Y-Ba-Cu-O superconductor in a usable form for magnet winding which requires the development of a cladding with demanding properties. Numerous recent efforts in cold forming Bi-based superconductor tapes have been successful because a silver tube can be used to constrain the ceramic material, which is sintered at much lower temperature than the Y-Ba-Cu-O. Typical high temperature metals which can be used to encase Y-Ba-Cu-O during sintering do not permit ready diffusion of oxygen as silver does. Recently, the full or partial recovery of superconductivity has been achieved in transition-metal- doped Y-Ba-Cu-O due to themore » partial-melt processing.« less
Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

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

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less
Refractory clad transient internal probe for magnetic field measurements in high temperature plasmas

NASA Astrophysics Data System (ADS)

Kim, Hyundae; Cellamare, Vincent; Jarboe, Thomas R.; Mattick, Arthur T.

2005-05-01

The transient internal probe (TIP) is a diagnostic for local internal field measurements in high temperature plasmas. A verdet material, which rotates the polarization angle of the laser light under magnetic fields, is launched into a plasma at about 1.8km/s. A linearly polarized Ar+ laser illuminates the probe in transit and the light retroreflected from the probe is analyzed to determine the local magnetic field profiles. The TIP has been used for magnetic field measurements on the helicity injected torus where electron temperature Te⩽80eV. In order to apply the TIP in higher temperature plasmas, refractory clad probes have been developed utilizing a sapphire tube, rear disc, and a MgO window on the front. The high melting points of these refractory materials should allow probe operation at plasma electron temperatures up to Te˜300eV. A retroreflecting probe has also been developed using "catseye" optics. The front window is replaced with a plano-convex MgO lens, and the back surface of the probe is aluminized. This approach reduces spurious polarization effects and provides refractory cladding for the probe entrance face. In-flight measurements of a static magnetic field demonstrate the ability of the clad probes to withstand gun-launch acceleration, and provide high accuracy measurements of magnetic field.
In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

DOE PAGES

Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; ...

2015-08-25

Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filteringmore » unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.« less
Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

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

Robb, Kevin R.

2015-08-01

Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramicmore » microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6 [10], and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.« less
DOE Office of Scientific and Technical Information (OSTI.GOV)

Ingham, J.G.

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

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

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.
Investigation of mechanical and microstructural properties of Zircaloy-4 under different experimental conditions

DOE PAGES

Silva, Chinthaka M.; Leonard, Keith J.; Van Abel, Eric; ...

2017-12-09

Here two types of Zircaloy-4 (alpha-annealed and beta-quenched) were investigated in their different forms. It was found that mechanical properties of Zircaloy-4 are affected significantly by welding and hydrogen-charging followed by neutron irradiation. Evaluation of microstructural properties of samples showed that these changes are mainly due to the formation of secondary phases such as hydrides—mostly along grain boundaries, dislocation channeling and their disruptions, and the increase in the type dislocation loops.
Investigation of mechanical and microstructural properties of Zircaloy-4 under different experimental conditions

NASA Astrophysics Data System (ADS)

Silva, Chinthaka M.; Leonard, Keith J.; Van Abel, Eric; Geringer, J. Wilna; Bryan, Chris D.

2018-02-01

Two types of Zircaloy-4 (alpha-annealed and beta-quenched) were investigated in their different forms. It was found that mechanical properties of Zircaloy-4 are affected significantly by welding and hydrogen-charging followed by neutron irradiation. Evaluation of microstructural properties of samples showed that these changes are mainly due to the formation of secondary phases such as hydrides-mostly along grain boundaries, dislocation channeling and their disruptions, and the increase in the type dislocation loops.

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)
2 Professors Rock Out Online to Study Fame--and Us

ERIC Educational Resources Information Center

Young, Jeffrey R.

2009-01-01

Most people who stumble across the YouTube video of the self-proclaimed rock star Gory Bateson singing to a scantily clad prostitute in Amsterdam's red-light district probably have no idea that the work is part of a research project--or that the man holding the guitar is a tenured professor. The video has attracted more than 12,000 views and won a…
DOE Office of Scientific and Technical Information (OSTI.GOV)

Ingham, J.G.

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

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

Field, Kevin G; Yamamoto, Yukinori; Pint, Bruce A

2016-01-01

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanicalmore » characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs.« less
Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

NASA Astrophysics Data System (ADS)

Metzger, Kathryn E.

Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels 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 or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it difficult to predict fuel-cladding mechanical behavior. This information is essential for designing accident tolerant fuel systems where ceramic claddings, like silicon carbide (SiC) are proposed. This research provides a model for both the thermal and irradiation creep behavior for U3Si2. This body of research is comprised of both experimental and modeling components. Characterization of the fuel microstructure includes; optical microscopy with pore and grain size analysis, helium pycnometry for density determination, mercury intrusion porosimetry, compositional analysis in the form of XRD, second phase identification using EDX, electrical resistance measurement via four point probe, determination of hardness and toughness through Vickers indentation testing, and determination of elastic properties using the impulse excitation method. Post-sintering grain size data allowed for the determination of grain boundary activation energy and diffusion coefficients, which were used to develop creep models. This was extended to lattice and irradiation enhanced diffusion in order to develop a U3Si2 creep model over thermal and irradiation creep regimes. In addition to the creep model, thermal and swelling behavior models for U3Si2 were implemented into the BISON fuel performance code. A series of simulations evaluated the performance and behavior of U3Si2 under typical light water reactor conditions with advanced SiC ceramic cladding. Simulation results show that fuel creep relieves stress in the ceramic cladding and postpones the. moment of fuel-clad contact. However, the stress reduction to the cladding is minimal because the fuel creep rate is low while the swelling rate is high. Future work should include the investigation of monolithic U3Si2 irradiation swelling since the current model relies upon the swelling data of U3Si2 particles in a metallic dispersion fuel. Additionally, planned thermal creep testing at the University of South Carolina can provide confirmation of the U3Si2 creep model contained herein.
Investigation of the weldability of iron-aluminum-chromium overlay coatings for corrosion protection in oxidizing/sulfidizing environments

NASA Astrophysics Data System (ADS)

Regina, Jonathan R.

The current study investigated the effect of chromium additions on the hydrogen cracking susceptibility of Fe-Al weld overlay claddings containing chromium additions. It was found that the weldability of FeAlCr claddings was a function of both the aluminum and chromium concentrations of the weld coatings. Weld overlay compositions that were not susceptible to hydrogen cracking were identified and the underlying mechanism behind the hydrogen cracking phenomenon was investigated further. It was concluded that the cracking behavior of the FeAlCr welds depended strongly on the microstructure of the weld fusion zone. Although it was found that the cracking susceptibility was influenced by the presence of Fe-Al intermetallic phases (namely Fe3 Al and FeAl), the cracking behavior of FeAlCr weld overlay claddings also depended on the size and distribution of carbide and oxide particles present within the weld structure. These particles acted as hydrogen trapping sites, which are areas where free hydrogen segregates and can no longer contribute to the hydrogen embrittlement of the metal. It was determined that in practical applications of these FeAlCr weld overlay coatings, carbon should be present within these welds to reduce the amount of hydrogen available for hydrogen cracking. Based on the weldability results of the FeAlCr weld claddings, coating compositions that were able to be deposited crack-free were used for long-term corrosion testing in a simulated low NOx environment. These alloys were compared to a Ni-based superalloy (622), which is commonly utilized as boiler tube coatings in power plant furnaces for corrosion protection. It was found that the FeAlCr alloys demonstrated superior corrosion resistance when compared to the Ni-based superalloy. Due to the excellent long-term corrosion behavior of FeAlCr weld overlays that were immune to hydrogen cracking, it was concluded that select FeAlCr weld overlay compositions would make excellent corrosion resistant coatings for boiler tubes located in low NOx burning environments.
Response of Cr and Cr-Al coatings on Zircaloy-2 to high temperature steam

NASA Astrophysics Data System (ADS)

Zhong, Weicheng; Mouche, Peter A.; Heuser, Brent J.

2018-01-01

The oxidation behavior of chromium (Cr) and chromium-aluminum (CrAl) coatings with various compositions deposited on Zircaloy-2 to 700 °C high-temperature steam (HTS) exposure has been investigated. CrAl coatings with higher Al compositions demonstrate lower oxidation weight gain. A layer of γ-alumina developed on the CrAl coatings with Al composition over 43 at%, while Al2O3 and Cr2O3 developed on CrAl coatings with Al composition below 33 at%. Oxidation of Zircaloy-2 substrate was inhibited by the 1um coatings to 20 h HTS exposure. Coating constituent elements diffused into the substrate and formed intermetallic phases with the Zircaloy substrate. Thicker layers of intermetallic phases developed on the coatings with higher Al composition. The intermetallic phases included Fe and Ni, indicating the dissolution of second phase particles (SPPs) during HTS exposure.
Proving the viability of manufacturing of multi-layer steel/vanadium alloy/steel composite tubes by numerical simulations and experiment

NASA Astrophysics Data System (ADS)

Nechaykina, T.; Nikulin, S.; Rozhnov, A.; Molotnikov, A.; Zavodchikov, S.; Estrin, Y.

2018-05-01

Vanadium alloys are promising structural materials for fuel cladding tubes for fast-neutron reactors. However, high solubility of oxygen and nitrogen in vanadium alloys at operating temperatures of 700 °C limits their application. In this work, we present a novel composite structure consisting of vanadium alloy V-4Ti-4Cr (provides high long-term strength of the material) and stainless steel Fe-0.2C-13Cr (as a corrosion resistant protective layer). It is produced by co-extrusion of these materials forming a three-layered tube. Finite element simulations were utilised to explore the influence of the various co-extrusion parameters on manufacturability of multi-layered tubes. Experimental verification of the numerical modelling was performed using co-extrusion with the process parameters suggested by the numerical simulations. Scanning electron microscopy and microhardness measurements revealed a defect-free diffusion layer at the interfaces between both materials indicating a good quality bonding for these co-extrusion conditions.
Purification of nuclear grade Zr scrap as the high purity dense Zr deposits from Zirlo scrap by electrorefining in LiF-KF-ZrF4 molten fluorides

NASA Astrophysics Data System (ADS)

Park, Kyoung Tae; Lee, Tae Hyuk; Jo, Nam Chan; Nersisyan, Hayk H.; Chun, Byong Sun; Lee, Hyuk Hee; Lee, Jong Hyeon

2013-05-01

Zirconium (Zr) has commonly been used as a cladding material of nuclear fuel. Moreover, it is regarded as the only material that can be used for nuclear fuel cladding because it has the lowest neutron capture cross section of any metal element and because it has high corrosion resistance and size stability. In this study, Hf-free Zr tubes (Zr-1Nb-1Sn-0.1Fe) were used as anode materials and electrorefining was performed in a LiF-KF eutectic 6 wt.% ZrF4 molten fluoride salt system. As a result of electrolysis, Zr scrap metal was recycled into pure Zr with low levels of impurities, and the size and density of the Zr deposit was controlled using applied current density.
Development of an Engineered Producet Storage Concept for the UREX+1 Combined Transuraqnic?Lanthanide Product Streams

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

Dr. Sean M. McDeavitt; Thomas J. Downar; Dr. Temitope A. Taiwo

2009-03-01

The U.S. Department of Energy is developing next generation processing methods to recycle uranium and transuranic (TRU) isotopes from spent nuclear fuel. The objective of the 3-year project described in this report was to develop near-term options for storing TRU oxides isolated through the uranium extraction (UREX+) process. More specifically, a Zircaloy matrix cermet was developed as a storage form for transuranics with the understanding that the cermet also has the ability to serve as a inert matrix fuel form for TRU burning after intermediate storage. The goals of this research projects were: 1) to develop the processing steps requiredmore » to transform the effluent TRU nitrate solutions and the spent Xircaloy cladding into a zireonium matrix cermet sotrage form; and 2) to evaluate the impact of phenomena that govern durability of the storage form, material processing, and TRU utiliztion in fast reactor fuel. This report represents a compilation of the results generated under this program. The information is presented as a brief technical narrative in the following sections with appended papers, presentations and academic theses to provide a detailed review of the project's accomplishments.« less
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.
High- and Low-Temperature Deformation Behavior of Different Orientation Hot-Rolled Annealed Zircaloy-4

NASA Astrophysics Data System (ADS)

Zong, Yingying; Gen, Qingfeng; Jiang, Hongwei; Shan, Debin; Guo, Bin

2018-03-01

In this paper, the hot-rolled annealed Zircaloy-4 samples with different orientation were subjected to uniaxial compression with a strain rate of 0.001 s-1 to obtain the stress-strain curves of different initial orientation samples at different temperatures. Electron backscatter diffraction (EBSD) technique and transmission electron microscope (TEM) technique were used to analyze the microstructures and textures of compressed samples. The mechanical properties and microstructural evolution of rolling directions (RD), transverse directions (TD) and normal directions (ND) were investigated under the conditions of - 150 °C low temperature, room temperature and 200 °C high temperature (simulated lunar temperature environment). The results show that the strength of Zircaloy-4 decreases with the increase in deformation temperature, and the strength in three orientations is ND > TD > RD. The deformation mechanism of hot-rolled annealed Zircaloy-4 with different orientation is different. In RD, { 10\\bar{1}0} < {a} > prismatic slip has the highest Schmid factor (SF), so it is most easy to activate the slip, followed by TD orientation, and ND orientation is the most difficult to activate. The deformed grains abide slip→twinning→slip rule, and the different orientation Zircaloy-4 deformation mechanisms mainly are the twinning coordinated with the slip.
Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

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

Ernst, Frank

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, andmore » fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.« less
METHOD OF MAKING WIRE FUEL ELEMENTS

DOEpatents

Zambrow, J.L.

1960-08-01

A method is given for making a nuclear reactor fuel element in the form of a uranium-bearing wire clad with zirconium. A uranium bar is enclosed in a zirconium sheath which is coated with an oxide of magnesium, beryllium, or zirconium. The sheathed bar is then placed in a steel tube and reduced to the desired diameter by swaging at 800 to 900 deg C, after which the steel and oxide are removed.
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.
Magnetic field sensor based on the magnetic-fluid-clad combined with singlemode-multimode-singlemode fiber and large core-offset splicing structure

NASA Astrophysics Data System (ADS)

Lv, Ri-qing; Qian, Jun-kai; Zhao, Yong

2018-03-01

A simple, compact optical fiber magnetic field sensor is proposed and experimentally demonstrated in this paper. It is based on the magnetic-fluid-clad combined with singlemode-multimode-singlemode fiber structure and large core-offset splicing structure. It was protected by a section of capillary tube and was sealed by UV glue. A sensing property study of the combined optical fiber structure and the proposed sensor were carried out. The experimental results show that the sensitivity of the refractive index of the optical fiber sensing structure is up to 156.63 nm/RIU and the magnetic field sensitivity of the proposed sensor is up to -97.24 pm/Oe in the range from 72.4 Oe to 297.8 Oe. The proposed sensor has several other advantages, such as simple structure, small size, easy fabrication and low cost.

Performance of OSC's initial Amtec generator design, and comparison with JPL's Europa Orbiter goals

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

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

1998-07-01

The procedure for the analysis (with overpotential correction) of multitube AMTEC (Alkali Metal Thermal-to-Electrical Conversion) cells described in Paper IECEC 98-243 was applied to a wide range of multicell radioisotope space power systems. System design options consisting of one or two generators, each with 2, 3, or 4 stacked GPHS (General Purpose Heat Source) modules, identical to those used on previous NASA missions, were analyzed and performance-mapped. The initial generators analyzed by OSC had 8 AMTEC cells on each end of the heat source stack, with five beta-alumina solid electrolyte (BASE) tubes per cell. The heat source and converters inmore » the Orbital generator designs are embedded in a thermal insulation system consisting of Min-K fibrous insulation surrounded by graded-length molybdenum multifoils. Detailed analyses in previous Orbital studies found that such an insulation system could reduce extraneous heat losses to about 10%. For the above design options, the present paper presents the system mass and performance (i.e., the EOM system efficiency and power output and the BOM evaporator and clad temperatures) for a wide range of heat inputs and load voltages, and compares the results with JPL's preliminary goals for the Europa Orbiter mission to be launched in November 2003. The analytical results showed that the initial 16-cell generator designs resulted in either excessive evaporator and clad temperatures and/or insufficient power outputs to meet the JPL-specified mission goals. The computed performance of modified OSC generators with different numbers of AMTEC cells, cell diameters, cell lengths, cell materials, BASE tube lengths, and number of tubes per cell are described in Paper IECEC.98.245 in these proceedings.« less
Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

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

Linton, Kory D.; Field, Kevin G.; Petrie, Christian M.

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. Tomore » address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).« less
Application of thin layer activation technique for surface wear studies in Zr based materials using charged particle induced nuclear reactions

NASA Astrophysics Data System (ADS)

Chowdhury, D. P.; Pal, Sujit; Parthasarathy, R.; Mathur, P. K.; Kohli, A. K.; Limaye, P. K.

1998-09-01

Thin layer activation (TLA) technique has been developed in Zr based alloy materials, e.g., zircaloy II, using 40 MeV α-particles from Variable Energy Cyclotron Centre at Calcutta. A brief description of the methodology of TLA technique is presented to determine the surface wear. The sensitivity of the measurement of surface wear in zircaloy material is found to be 0.22±0.05 μm. The surface wear is determined by TLA technique in zircaloy material which is used in pressurised heavy water reactor and the values have been compared with that obtained by conventional technique for the analytical validation of the TLA technique.
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.
Test Results of Heat Exchanger Cleaning in Support of Ocean Thermal Energy Conversion.

DTIC Science & Technology

1980-12-01

tests evaluated the performance of three in-situ cleaning techniques in two potential heat exchanger materials ...1-6. 41Mann, M. J., 1979, "Possible Cu-Ni-Clad Steel Material and Abrasive Slurry Cleaning System for Plate-Fin-Type OTEC Heat Exchangers ," in...of a Shell-less Folded Aluminum Tube, OTEC Heat Exchanger ," Proceedings of the Sixth OTEC Conference, Washington, DC, June 19-22, 1978, pp 12.8-1
77 FR 13156 - Carolina Power & Light Company; Shearon Harris Nuclear Power Plant, Unit 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-05

... percent) and niobium (~1 percent). The elimination of tin has resulted in superior corrosion resistance and reduced irradiation-induced growth relative to both standard zircaloy (1.7 percent tin) and low-tin zircaloy (1.2 percent tin). The addition of niobium increases ductility, which is desirable to...
76 FR 68512 - Carolina Power & Light Company; H. B. Robinson Steam Electric Plant, Unit 2; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-04

... (~1 percent). The elimination of tin has resulted in superior corrosion resistance and reduced irradiation-induced growth relative to both standard zircaloy (1.7 percent tin) and low-tin zircaloy (1.2 percent tin). The addition of niobium increases ductility, which is desirable to avoid brittle failures...
Brazing characteristics of a Zr-Ti-Cu-Fe eutectic alloy filler metal for Zircaloy-4

NASA Astrophysics Data System (ADS)

Lee, Jung G.; Lim, C. H.; Kim, K. H.; Park, S. S.; Lee, M. K.; Rhee, C. K.

2013-10-01

A Zr-Ti-Cu-Fe quaternary eutectic alloy was employed as a new Be-free brazing filler metal for Zircaloy-4 to supersede physically vapor-deposited Be coatings used conventionally with several disadvantages. The quaternary eutectic composition of Zr58Ti16Cu10Fe16 (at.%) showing a low melting temperature range from 832 °C to 853 °C was designed by a partial substitution of Zr with Ti based on a Zr-Cu-Fe ternary eutectic system. By applying an alloy ribbon with the determined composition, a highly reliable joint was obtained with a homogeneous formation of predominantly grown α-Zr phases owing to a complete isothermal solidification, exhibiting strength higher than that of Zircaloy-4. The homogenization of the joint was rate-controlled by the diffusion of the filler elements (Ti, Cu, and Fe) into the Zircaloy-4 base metal, and the detrimental segregation of the Zr2Fe phase in the central zone was completely eliminated by an isothermal holding at a brazing temperature of 920 °C for 10 min.
Linear Friction Welding of Dissimilar Materials 316L Stainless Steel to Zircaloy-4

NASA Astrophysics Data System (ADS)

Wanjara, P.; Naik, B. S.; Yang, Q.; Cao, X.; Gholipour, J.; Chen, D. L.

2018-02-01

In the nuclear industry, there are a number of applications where the transition of stainless steel to Zircaloy is of technological importance. However, due to the differences in their properties there are considerable challenges associated with developing a joining process that will sufficiently limit the heat input and welding time—so as to minimize the extent of interaction at the joint interface and the resulting formation of intermetallic compounds—but still render a functional metallurgical bond between these two alloys. As such, linear friction welding, a solid-state joining technology, was selected in the present study to assess the feasibility of welding 316L stainless steel to Zircaloy-4. The dissimilar alloy welds were examined to evaluate their microstructural characteristics, microhardness evolution across the joint interface, static tensile properties, and fatigue behavior. Microstructural observations revealed a central intermixed region and, on the Zircaloy-4 side, dynamically recrystallized and thermomechanically affected zones were present. By contrast, deformation on the 316L stainless steel side was limited. In the intermixed region a drastic change in the composition was observed along with a local increase in hardness, which was attributed to the presence of intermetallic compounds, such as FeZr3 and Cr2Zr. The average yield (316 MPa) and ultimate tensile (421 MPa) strengths met the minimum strength properties of Zircaloy-4, but the elongation was relatively low ( 2 pct). The tensile and fatigue fracture of the welds always occurred at the interface in the mode of partial cohesive failure.
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
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.
CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT ...

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

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT BREAKS IN CLADDING OF FUEL ELEMENTS. COUNT-RATE METER IN TOP PANEL INDICATES AMOUNT OF RADIOACTIVITY. LOWER PANELS SUPPLY POWER AND AMPLIFICATION OF SIGNALS GENERATED BY SCINTILLATION COUNTER/PHOTOMULTIPLIER TUBE COMBINATION IN RESPONSE TO RADIOACTIVITY IN A SAMPLE OF THE COOLING WATER. INL NEGATIVE NO. 56-771. Jack L. Anderson, Photographer, 3/15/1956. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID
Analysis of Ignition Testing on K-West Basin Fuel

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

J. Abrefah; F.H. Huang; W.M. Gerry

Approximately 2100 metric tons of spent nuclear fuel (SNF) discharged from the N-Reactor have been stored underwater at the K-Basins in the 100 Area of the Hanford Site. The spent fuel has been stored in the K-East Basin since 1975 and in the K-West Basin since 1981. Some of the SNF elements in these basins have corroded because of various breaches in the Zircaloy cladding that occurred during fuel discharge operations and/or subsequent handling and storage in the basins. Consequently, radioactive material in the fuel has been released into the basin water, and water has leaked from the K-East Basinmore » into the soil below. To protect the Columbia River, which is only 380 m from the basins, the SNF is scheduled to be removed and transported for interim dry storage in the 200 East Area, in the central portion of the Site. However, before being shipped, the corroded fuel elements will be loaded into Multi-Canister OverPacks and conditioned. The conditioning process will be selected based on the Integrated Process Strategy (IPS) (WHC 1995), which was prepared on the basis of the dry storage concept developed by the Independent Technical Assessment (ITA) team (ITA 1994).« less
303-K Storage Facility closure plan. Revision 2

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

Not Available

1993-12-15

Recyclable scrap uranium with zircaloy-2 and copper silicon alloy, uranium-titanium alloy, beryllium/zircaloy-2 alloy, and zircaloy-2 chips and fines were secured in concrete billets (7.5-gallon containers) in the 303-K Storage Facility, located in the 300 Area. The beryllium/zircaloy-2 alloy and zircaloy-2 chips and fines are designated as mixed waste with the characteristic of ignitability. The concretion process reduced the ignitability of the fines and chips for safe storage and shipment. This process has been discontinued and the 303-K Storage Facility is now undergoing closure as defined in the Resource Conservation and Recovery Act (RCRA) of 1976 and the Washington Administrative Codemore » (WAC) Dangerous Waste Regulations, WAC 173-303-040. This closure plan presents a description of the 303-K Storage Facility, the history of materials and waste managed, and the procedures that will be followed to close the 303-K Storage Facility. The 303-K Storage Facility is located within the 300-FF-3 (source) and 300-FF-5 (groundwater) operable units, as designated in the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology et al. 1992). Contamination in the operable units 300-FF-3 and 300-FF-5 is scheduled to be addressed through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) of 1980 remedial action process. Therefore, all soil remedial action at the 304 Facility will be conducted as part of the CERCLA remedial action of operable units 300-FF-3 and 300-FF-5.« less
TEM/STEM study of Zircaloy-2 with protective FeAl(Cr) layers under simulated BWR environment and high-temperature steam exposure

NASA Astrophysics Data System (ADS)

Park, Donghee; Mouche, Peter A.; Zhong, Weicheng; Mandapaka, Kiran K.; Was, Gary S.; Heuser, Brent J.

2018-04-01

FeAl(Cr) thin-film depositions on Zircaloy-2 were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) with respect to oxidation behavior under simulated boiling water reactor (BWR) conditions and high-temperature steam. Columnar grains of FeAl with Cr in solid solution were formed on Zircaloy-2 coupons using magnetron sputtering. NiFe2O4 precipitates on the surface of the FeAl(Cr) coatings were observed after the sample was exposed to the simulated BWR environment. High-temperature steam exposure resulted in grain growth and consumption of the FeAl(Cr) layer, but no delamination at the interface. Outward Al diffusion from the FeAl(Cr) layer occurred during high-temperature steam exposure (700 °C for 3.6 h) to form a 100-nm-thick alumina oxide layer, which was effective in mitigating oxidation of the Zircaloy-2 coupons. Zr intermetallic precipitates formed near the FeAl(Cr) layer due to the inward diffusion of Fe and Al. The counterflow of vacancies in response to the Al and Fe diffusion led to porosity within the FeAl(Cr) layer.
LIGHT WATER MODERATED NEUTRONIC REACTOR

DOEpatents

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

1957-09-17

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

NASA Astrophysics Data System (ADS)

Stockdale, Andrew

The use of low NOx boilers in coal fired power plants has resulted in sulfidizing corrosive conditions within the boilers and a reduction in the service lifetime of the waterwall tubes. As a solution to this problem, Ni-based weld overlays are used to provide the necessary corrosion resistance however; they are susceptible to corrosion fatigue. There are several metallurgical factors which give rise to corrosion fatigue that are associated with the localized melting and solidification of the weld overlay process. Coextruded coatings offer the potential for improved corrosion fatigue resistance since coextrusion is a solid state coating process. The corrosion and corrosion fatigue behavior of alloy 622 weld overlays and coextruded claddings was investigated using a Gleeble thermo-mechanical simulator retrofitted with a retort. The experiments were conducted at a constant temperature of 600°C using a simulated combustion gas of N2-10%CO-5%CO2-0.12%H 2S. An alternating stress profile was used with a minimum tensile stress of 0 MPa and a maximum tensile stress of 300 MPa (ten minute fatigue cycles). The results have demonstrated that the Gleeble can be used to successfully simulate the known corrosion fatigue cracking mechanism of Ni-based weld overlays in service. Multilayer corrosion scales developed on each of the claddings that consisted of inner and outer corrosion layers. The scales formed by the outward diffusion of cations and the inward diffusion of sulfur and oxygen anions. The corrosion fatigue behavior was influenced by the surface finish and the crack interactions. The initiation of a large number of corrosion fatigue cracks was not necessarily detrimental to the corrosion fatigue resistance. Finally, the as-received coextruded cladding exhibited the best corrosion fatigue resistance.
State Environmental Policy Act (SEPA) environmental checklist forms for 304 Concretion Facility Closure Plan. Revision 2

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

Not Available

The 300 Area of the Hanford Site contains reactor fuel manufacturing facilities and several research and development laboratories. Recyclable scrap uranium with zircaloy-2 and copper silicon alloy, uranium-titanium alloy, beryllium/zircaloy-2 alloy, and zircaloy-2 chips and fines were secured in concrete billets (7.5-gallon containers) in the 304 Facility, located in the 300 Area. The beryllium/zircaloy-2 alloy and zircaloy-2 chips and fines are designated as mixed waste with the characteristic of ignitability. The concretion process reduced the ignitability of the fines and chips for safe storage and shipment. This process has been discontinued and the 304 Facility is now undergoing closure asmore » defined in the Resource Conservation and Recovery Act (RCRA) of 1976 and the Washington Administrative Code (WAC) Dangerous Waste Regulations, WAC 173-303-040. This closure plan presents a description of the 304 Facility, the history of materials and waste managed, and the procedures that will be followed to close the 304 Facility. The 304 Facility is located within the 300-FF-3 (source) and 300-FF-5 (groundwater) operable units, as designated in the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology et al. 1992). Contamination in the operable units 300-FF-3 and 300-FF-5 is scheduled to be addressed through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) of 1980 remedial action process. Therefore, all soil remedial action at the 304 Facility will be conducted as part of the CERCLA remedial action of operable units 300-FF-3 and 300-FF-5.« less
The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

NASA Astrophysics Data System (ADS)

Cockeram, B. V.; Kammenzind, B. F.

2018-02-01

Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

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
Recording of Electric Signal Passing Through a Pylon in Direct Skeletal Attachment of Leg Prostheses with Neuromuscular Control

PubMed Central

Pitkin, M.; Cassidy, C.; Muppavarapu, R.; Edell, David

2012-01-01

Direct recordings were made of electrical signals emanating from the muscles in a rabbit’s residuum. The signals were transmitted via wires attached on one end to the muscles, and on the other to an external recording system. The cable was held in a titanium tube inside a pylon that had been transcutaneously implanted into the residuum’s bone. The tube was surrounded by porous titanium cladding to enhance its bond with the bone and with the skin of the residuum. This study was the first known attempt to merge the technology of direct skeletal attachment of limb prostheses with the technology of neuromuscular control of prostheses, providing a safe and reliable passage of the electrical signal from the muscles inside the residuum to the outside recording system. PMID:22345523
THE DETERMINATION OF BORON IN ZIRCALOY

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

Freegarde, M.; Cartwright, J.

1962-03-01

An account is given of the development of a simple and reliable procedure for determining boron in Zircaloy at the parts per million level. The sample is dissolved in a mixture of bromine and methanol, and the boron is separated by distillation and determined as its rosocyanin complex with curcumin. The reproducibility of the method is characterized by a standard deviation of 0.03 ppm at the 0.3 ppm level. (auth)
Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

NASA Astrophysics Data System (ADS)

Lee, Jung Gu; Lee, Gyoung-Ja; Park, Jin-Ju; Lee, Min-Ku

2017-05-01

The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments.
Modified Powder-in-Tube Technique Based on the Consolidation Processing of Powder Materials for Fabricating Specialty Optical Fibers

PubMed Central

Auguste, Jean-Louis; Humbert, Georges; Leparmentier, Stéphanie; Kudinova, Maryna; Martin, Pierre-Olivier; Delaizir, Gaëlle; Schuster, Kay; Litzkendorf, Doris

2014-01-01

The objective of this paper is to demonstrate the interest of a consolidation process associated with the powder-in-tube technique in order to fabricate a long length of specialty optical fibers. This so-called Modified Powder-in-Tube (MPIT) process is very flexible and paves the way to multimaterial optical fiber fabrications with different core and cladding glassy materials. Another feature of this technique lies in the sintering of the preform under reducing or oxidizing atmosphere. The fabrication of such optical fibers implies different constraints that we have to deal with, namely chemical species diffusion or mechanical stress due to the mismatches between thermal expansion coefficients and working temperatures of the fiber materials. This paper focuses on preliminary results obtained with a lanthano-aluminosilicate glass used as the core material for the fabrication of all-glass fibers or specialty Photonic Crystal Fibers (PCFs). To complete the panel of original microstructures now available by the MPIT technique, we also present several optical fibers in which metallic particles or microwires are included into a silica-based matrix. PMID:28788176
Glass-ceramic optical fiber containing Ba2TiSi2O8 nanocrystals for frequency conversion of lasers

NASA Astrophysics Data System (ADS)

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

2017-03-01

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

PubMed Central

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

2017-01-01

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

PubMed

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

2017-03-30

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

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

Soulami, Ayoub; Burkes, Douglas E.; Joshi, Vineet V.

This study used a finite element code, LSDYNA, as a predictive tool to optimize the rolling process. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel were conducted following two different schedules. Model predictions of the roll-separation force and roll pack thicknesses at different stages of the rolling process were compared with experimental measurements. The study reported here discussed various attributes of the rolled coupons revealed by the model (e.g., waviness and thickness non-uniformity like dog boning). To investigate the influence of the cladding material on these rolling defects, other cases were simulated: hot rolling with alternative can materials, namely, 304 stainless steel and Zircaloy-2, and bare-rolling.
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).
Effect of nitrogen on tensile properties and structures of T-111 (tantalum, 8 percent tungsten, 2 percent hafnium) tubing

NASA Technical Reports Server (NTRS)

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

1973-01-01

The effect of controlled nitrogen additions was evaluated on the mechanical properties of T-111 (Ta-8W-2Hf) fuel pin cladding material proposed for use in a lithium-cooled nuclear reactor concept. Additions of 80 to 1125 ppm nitrogen resulted in increased strengthening of T-111 tubular section test specimens at temperatures of 25 to 1200 C. Homogeneous distributions of up to 500 ppm nitrogen did not seriously decrease tensile ductility. Both single and two-phase microstructures, with hafnium nitride as the second phase, were evaluated in this study.
EXAMINATION OF Zr AND Ti RECOMBINER LOOP SPECIMENS

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

Rittenhouse, P.L.

1958-12-19

Cold-worked specimens of iodide zirconium, Zircaloy-2, iodide titanium, and A-55 titanium were tested in a high-pressure recombiner loop in an attempt to duplicate anomalous results obtained in a prior recombiner loop. Hydrogen analyses and metallographic examinations were made on all specimens. The titanium materials and Zircaloy-2 picked up major amounts of hydrogen in the cell section. None of the materials tested showed appreciable hydrogen absorption in the recombiner section. Complete recrystallization occurred in all cell specimens while only Zircaloy-2, of the recombiner specimens, showed any degree of recrystallization. No explanation for this behavior can be given. A survnnary of themore » data obtained in previous recombiner loops is compared with the results of this loop. Conclusions were based on the results of three recombiner loops. Primarlly because of the hydrogen absorption data obtained in all three recombiner loops it is recommended that the zirconium and titunium materials tested not be used in environments similar to those encountered in high pressure recombiner loops. (auth)« less
Room temperature mechanical properties of electron beam welded zircaloy-4 sheet

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

Parga, C. J.; Rooyen, I. J.; Coryell, B. D.

Room temperature mechanical properties of electron beam welded and plain Zircaloy-4 sheet (1.6mm thick) have been measured and compared. Various welding parameters were utilized to join sheet material. Electron beam welded specimens and as-received sheet specimens show comparable mechanical properties. Zr-4 sheet displays anisotropy; tensile properties measured for transverse display higher elastic modulus, yield strength, reduction of area and slightly lower ductility than for the longitudinal (rolling direction). Higher welding power increases the alloy’s hardness, elastic modulus and yield strength, with a corresponding decrease in tensile strength and ductility. The hardness measured at weld is comparable to the parent metalmore » hardness. Hardness at heat-affected-zone is slightly higher. Electron microscopic examination shows distinct microstructure morphology and grain size at the weld zone, HAZ and parent metal. A correlation between welding parameters, mechanical properties and microstructural features was established for electron beam welded Zircaloy-4 sheet material.« less
Room temperature mechanical properties of electron beam welded zircaloy-4 sheet

DOE PAGES

Parga, C. J.; Rooyen, I. J.; Coryell, B. D.; ...

2017-11-04

Room temperature mechanical properties of electron beam welded and plain Zircaloy-4 sheet (1.6mm thick) have been measured and compared. Various welding parameters were utilized to join sheet material. Electron beam welded specimens and as-received sheet specimens show comparable mechanical properties. Zr-4 sheet displays anisotropy; tensile properties measured for transverse display higher elastic modulus, yield strength, reduction of area and slightly lower ductility than for the longitudinal (rolling direction). Higher welding power increases the alloy’s hardness, elastic modulus and yield strength, with a corresponding decrease in tensile strength and ductility. The hardness measured at weld is comparable to the parent metalmore » hardness. Hardness at heat-affected-zone is slightly higher. Electron microscopic examination shows distinct microstructure morphology and grain size at the weld zone, HAZ and parent metal. A correlation between welding parameters, mechanical properties and microstructural features was established for electron beam welded Zircaloy-4 sheet material.« less
Review of PWR fuel rod waterside corrosion behavior

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

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

Waterside corrosion of Zircaloy has generally not been a problem under normal PWR operating conditions, although some instances of accelerated corrosion have been reported. However, an incentive exists to extend the average fuel rod discharge burnups to about 50,000 MWd/MTU. To minimize corrosion at these extended burnups, the factors which influence Zircaloy corrosion need to be better understood. A data base of Zircaloy corrosion behavior under PWR operating conditions has been established. The data are compiled previously published reports as well as from new Kraftwerk Union examinations. A non-destructive eddy-current technique is used to measure the oxide layer thickness onmore » fuel rods. Comparisons of measuremnts made using this eddy-current technique with those made by usual metallographic methods indicate good agreement. The data were evaluated by defining a fitting factor F which describes the increase in corrosion rate observed in-reactor over that observed from measurements of ex-reactor corrosion coupons.« less
Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

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

Dryepondt, Sebastien N.; Hoelzer, David T.; Pint, Bruce A.

2015-09-18

FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO 2 or ZrO 2. Themore » new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.« less
Optimized Gen-II FeCrAl cladding production in large quantity for campaign testing

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

Yamamoto, Yukinori; Sun, Zhiqian; Pint, Bruce A.

2016-06-03

There are two major objectives in this report; (1) to optimize microstructure control of ATF FeCrAl alloys during tube drawing processes, and (2) to provide an update on the progress of ATF FeCrAl tube production via commercial manufacturers. Experimental efforts have been made to optimize the process parameters balancing the tube fabricability, especially for tube drawing processes, and microstructure control of the final tube products. Lab-scale sheet materials of Gen II FeCrAl alloys (Mo-containing and Nb-containing FeCrAl alloys) were used in the study, combined with a stepwise warm-rolling process and intermediate annealing, aiming to simulate the tube drawing process inmore » a commercial tube manufacturer. The intermediate annealing at 650ºC for 1h was suggested for the tube-drawing process of Mo-containing FeCrAl alloys because it successfully softened the material by recovering the work hardening introduced through the rolling step, without inducing grain coarsening due to recrystallization. The final tube product is expected to have stabilized deformed microstructure providing the improved tensile properties with sufficient ductility. Optimization efforts on Nb-containing FeCrAl alloys focused on the effect of alloying additions and annealing conditions on the stability of deformed microstructure. Relationships between the second-phase precipitates (Fe 2Nb-Laves phase) and microstructure stability are discussed. FeCrAl tube production through commercial tube manufacturers is currently in progress. Three different manufacturers, Century Tubes, Inc. (CTI), Rhenium Alloys, Inc. (RAI), and Superior Tube Company, Inc. (STC), are providing capabilities for cold-drawing, warm-drawing, and HPTR cold-pilgering, respectively. The first two companies are currently working on large quantity tube production (expected 250 ft length) of Gen I model FeCrAl alloy (B136Y3, at CTI) and Gen II (C35M4, at RAI), with the process parameters obtained from the experimental efforts. The expected delivery dates are at the end of July, 2016, and the middle of June, 2016, respectively. Tube production at STC would be the first attempt to apply cold-pilgering to the FeCrAl alloys. Communication has been initiated, and the materials have been machined for the cold-pilgering process.« less
Writing Bragg Gratings in Multicore Fibers.

PubMed

Lindley, Emma Y; Min, Seong-Sik; Leon-Saval, Sergio G; Cvetojevic, Nick; Lawrence, Jon; Ellis, Simon C; Bland-Hawthorn, Joss

2016-04-20

Fiber Bragg gratings in multicore fibers can be used as compact and robust filters in astronomical and other research and commercial applications. Strong suppression at a single wavelength requires that all cores have matching transmission profiles. These gratings cannot be inscribed using the same method as for single-core fibers because the curved surface of the cladding acts as a lens, focusing the incoming UV laser beam and causing variations in exposure between cores. Therefore we use an additional optical element to ensure that the beam shape does not change while passing through the cross-section of the multicore fiber. This consists of a glass capillary tube which has been polished flat on one side, which is then placed over the section of the fiber to be inscribed. The laser beam enters the fiber through the flat surface of the capillary tube and hence maintains its original dimensions. This paper demonstrates the improvements in core-to-core uniformity for a 7-core fiber using this method. The technique can be generalized to larger multicore fibers.
Writing Bragg Gratings in Multicore Fibers

PubMed Central

Lindley, Emma Y.; Min, Seong-sik; Leon-Saval, Sergio G.; Cvetojevic, Nick; Lawrence, Jon; Ellis, Simon C.; Bland-Hawthorn, Joss

2016-01-01

Fiber Bragg gratings in multicore fibers can be used as compact and robust filters in astronomical and other research and commercial applications. Strong suppression at a single wavelength requires that all cores have matching transmission profiles. These gratings cannot be inscribed using the same method as for single-core fibers because the curved surface of the cladding acts as a lens, focusing the incoming UV laser beam and causing variations in exposure between cores. Therefore we use an additional optical element to ensure that the beam shape does not change while passing through the cross-section of the multicore fiber. This consists of a glass capillary tube which has been polished flat on one side, which is then placed over the section of the fiber to be inscribed. The laser beam enters the fiber through the flat surface of the capillary tube and hence maintains its original dimensions. This paper demonstrates the improvements in core-to-core uniformity for a 7-core fiber using this method. The technique can be generalized to larger multicore fibers. PMID:27167576
Aluminium-stabilized magnesium diboride—a new light-weight superconductor

NASA Astrophysics Data System (ADS)

Dou, S. X.; Collings, E. W.; Shcherbakova, O.; Shcherbakov, A.

2006-04-01

As a stabilizer for low-temperature superconductors, Al has found limited use due to the metallurgical difficulty and low melting point of Al. However, now that the processing of MgB2 wires at 600 °C has been demonstrated, all of the advantages of Al stabilization can be realized. With Al stabilization in mind, we describe in situ powder-in-tube 'low temperature processing' of mixed Mg+B powders in an Al tube lined with a protective Fe barrier. Reaction heat treatment at 600 °C, for up to 3 h, led to complete MgB2 formation; furthermore, no reaction between the Fe barrier and the Al sheath took place at 620 °C. The Fe/Al clad wires showed the same magnetic and electrical properties as those with an all-Fe sheath. The MgB2/Fe/Al conductor, mainly made up of low-density components, will be advantageous for airborne, aerospace, and other applications where weight is important.

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

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

Tentner, A.M.

1994-03-01

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

DOEpatents

Wigner, E.P.

1957-09-17

A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.
Study of a ;hot; particle with a matrix of U-bearing metallic Zr: Clue to supercriticality during the Chernobyl nuclear accident

NASA Astrophysics Data System (ADS)

Pöml, P.; Burakov, B.

2017-05-01

This paper is dedicated to the 30th anniversary of the severe nuclear accident that occurred at the Chernobyl NPP on 26 April 1986. A detailed study on a Chernobyl "hot" particle collected from contaminated soil was performed. Optical and electron microscopy, as well as quantitative x-ray microbeam analysis methods were used to determine the properties of the sample. The results show that the particle (≈ 240 x 165 μm) consists of a metallic Zr matrix containing 2-3 wt. % U and bearing veins of an U,Nb admixture. The metallic Zr matrix contains two phases with different amounts of O with the atomic proportions (U,Zr,Nb)0.73O0.27 and (U,Zr,Nb)0.61O0.39. The results confirm the interaction between UO2 fuel and zircaloy cladding in the reactor core. To explain the process of formation of the particle, its properties are compared to laboratory experiments. Because of the metallic nature of the particle it is concluded that it must have formed during a very high temperature (> 2400∘C) process that lasted for only a very short time (few microseconds or less); otherwise the particle should have been oxidised. Such a rapid very high temperature process indicates that at least part of the reactor core could have been supercritical prior to an explosion as it was previously suggested in the literature.
Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

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

Jacobsen, George; Back, Christina

2015-10-30

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called themore » endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.« less
Cyclic softening in annealed Zircaloy-2: Role of edge dislocation dipoles and vacancies

NASA Astrophysics Data System (ADS)

Sudhakar Rao, G.; Singh, S. R.; Krsjak, Vladimir; Singh, Vakil

2018-04-01

The mechanism of cyclic softening in annealed Zircaloy-2 at low strain amplitudes under strain controlled fatigue at room temperature is rationalized. The unusual softening due to continuous decrease in the phenomenological friction stress is found to be associated with decrease in the resistance against movement of dislocations because of the formation and easy glide of pure edge dislocation dipoles and consequent decrease in friction stress from reduction in the shear modulus. Positron annihilation spectroscopy data strongly support the increase in edge dislocation density containing jogs, from increased positron trapping and increase in annihilation lifetime.
Quantitative analysis of deuterium in zircaloy using double-pulse laser-induced breakdown spectrometry (LIBS) and helium gas plasma without a sample chamber.

PubMed

Suyanto, H; Lie, Z S; Niki, H; Kagawa, K; Fukumoto, K; Rinda, Hedwig; Abdulmadjid, S N; Marpaung, A M; Pardede, M; Suliyanti, M M; Hidayah, A N; Jobiliong, E; Lie, T J; Tjia, M O; Kurniawan, K H

2012-03-06

A crucial safety measure to be strictly observed in the operation of heavy-water nuclear power plants is the mandatory regular inspection of the concentration of deuterium penetrated into the zircaloy fuel vessels. The existing standard method requires a tedious, destructive, and costly sample preparation process involving the removal of the remaining fuel in the vessel and melting away part of the zircaloy pipe. An alternative method of orthogonal dual-pulse laser-induced breakdown spectrometry (LIBS) is proposed by employing flowing atmospheric helium gas without the use of a sample chamber. The special setup of ps and ns laser systems, operated for the separate ablation of the sample target and the generation of helium gas plasma, respectively, with properly controlled relative timing, has succeeded in producing the desired sharp D I 656.10 nm emission line with effective suppression of the interfering H I 656.28 nm emission by operating the ps ablation laser at very low output energy of 26 mJ and 1 μs ahead of the helium plasma generation. Under this optimal experimental condition, a linear calibration line is attained with practically zero intercept and a 20 μg/g detection limit for D analysis of zircaloy sample while creating a crater only 10 μm in diameter. Therefore, this method promises its potential application for the practical, in situ, and virtually nondestructive quantitative microarea analysis of D, thereby supporting the more-efficient operation and maintenance of heavy-water nuclear power plants. Furthermore, it will also meet the anticipated needs of future nuclear fusion power plants, as well as other important fields of application in the foreseeable future.
Corrosion behavior and oxide properties of Zr 1.1 wt%Nb 0.05 wt%Cu alloy

NASA Astrophysics Data System (ADS)

Park, Jeong-Yong; Choi, Byung-Kwon; Yoo, Seung Jo; Jeong, Yong Hwan

2006-12-01

The corrosion behavior and oxide properties of Zr-1.1 wt%Nb-0.05 wt%Cu (ZrNbCu) and Zircaloy-4 have been investigated. The corrosion rate of the ZrNbCu alloy was much lower than that of the Zirclaoy-4 in the 360 °C water and 360 °C PWR-simulating loop condition without a neutron flux and it was increased with an increase of the final annealing temperature from 470 °C to 570 °C. TEM observations revealed that the precipitates in the ZrNbCu were β-Nb and ZrNbFe-precipitate with β-Nb being more frequently observed and that the precipitates were more finely distributed in the ZrNbCu alloy. It was also observed that the oxides of the ZrNbCu and Zircaloy-4 consisted of two and seven layers, respectively, after 1000 days in the PWR-simulating loop condition and that the thickness of a fully-developed layer was higher in the ZrNbCu than in the Zircaloy-4. It was also found that the β-Nb in ZrNbCu was oxidized more slowly when compared to the Zr(Fe, Cr) 2 in Zirclaoy-4 when the precipitates in the oxide were observed by TEM. Cracks were observed in the vicinity of the oxidized Zr(Fe, Cr) 2, while no cracks were formed near β-Nb which had retained a metallic state. From the results obtained, it is suggested that the oxide formed on the ZrNbCu has a more protective nature against a corrosion when compared to that of the Zircaloy-4.
Characterization of Tubing from Advanced ODS alloy (FCRD-NFA1)

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

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

2016-09-20

Fabrication methods are being developed and tested for producing fuel clad tubing of the advanced ODS 14YWT and FCRD-NFA1 ferritic alloys. Three fabrication methods were based on plastically deforming a machined thick-wall tube sample of the ODS alloys by pilgering, hydrostatic extrusion or drawing to decrease the outer diameter and wall thickness and increase the length of the final tube. The fourth fabrication method consisted of the additive manufacturing approach involving solid-state spray deposition (SSSD) of ball milled and annealed powder of 14YWT for producing thin-wall tubes. Of the four fabrication methods, two methods were successful at producing tubing formore » further characterization: production of tubing by high-velocity oxy-fuel spray forming and production of tubing using high-temperature hydrostatic extrusion. The characterization described shows through neutron diffraction the texture produced during extrusion while maintaining the beneficial oxide dispersion. In this research, the parameters for innovative thermal spray deposition and hot extrusion processing methods have been developed to produce the final nanostructured ferritic alloy (NFA) tubes having approximately 0.5 mm wall thickness. Effect of different processing routes on texture and grain boundary characteristics has been investigated. It was found that hydrostatic extrusion results in combination of plane strain and shear deformations which generate rolling textures of α- and γ-fibers on {001}<110> and {111}<110> together with a shear texture of ζ-fiber on {011}<211> and {011}<011>. On the other hand, multi-step plane strain deformation in cross directions leads to a strong rolling textures of θ- and ε-fiber on {001}<110> together with weak γ-fiber on {111}<112>. Even though the amount of the equivalent strain is similar, shear deformation leads to much lower texture indexes compared to the plane strain deformations. Moreover, while 50% of hot rolling brings about a large number of high-angle grain boundaries (HAB), 44% of shear deformation results in large amount of low-angle boundaries (LAB) showing the incomplete recrystallization.« less
Cryogenic Design of a Large Superconducting Magnet for Astro-particle Shielding on Deep Space Travel Missions

NASA Astrophysics Data System (ADS)

Bruce, Romain; Baudouy, Bertrand

The Space Radiation Superconducting Shield (SR2S) European project aims at studying a large superconducting toroid magnet to protect the human habitat from the ionizing radiations coming from Galactic Cosmic Ray during long term missions in deep space. Titanium clad MgB2 conductor is used to afford a bending power greater than 5 T.m at 10 K. A specific cryogenic design is needed to cool down this 10 m long and 12.8 m in diameter magnet. A passive cooling system, using a V-groove sunshield, is considered to reduce the heat flux coming from the Sun or Mars. An active configuration, using pulse tube cryocoolers, will be linked to the 80 K thermal screen intercepting most of the heat fluxes coming from the human habitat. The toroid magnet will be connected also to cryocoolers to absorb the few watts reaching its surface. Two kinds of thermal link are being considered to absorb the heat on the 80 K thermal screen. The first one is active, with a pump circulating helium gas in a network of exchange tubes. The second one is passive using long cryogenic pulse heat pipe (PHP) with the evaporator on the surface of the thermal screen and the condenser attached to the pulse tube.
In-reactor oxidation of zircaloy-4 under low water vapor pressures

NASA Astrophysics Data System (ADS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.
Anisotropic hydrogen diffusion in α-Zr and Zircaloy predicted by accelerated kinetic Monte Carlo simulations

NASA Astrophysics Data System (ADS)

Zhang, Yongfeng; Jiang, Chao; Bai, Xianming

2017-01-01

This report presents an accelerated kinetic Monte Carlo (KMC) method to compute the diffusivity of hydrogen in hcp metals and alloys, considering both thermally activated hopping and quantum tunneling. The acceleration is achieved by replacing regular KMC jumps in trapping energy basins formed by neighboring tetrahedral interstitial sites, with analytical solutions for basin exiting time and probability. Parameterized by density functional theory (DFT) calculations, the accelerated KMC method is shown to be capable of efficiently calculating hydrogen diffusivity in α-Zr and Zircaloy, without altering the kinetics of long-range diffusion. Above room temperature, hydrogen diffusion in α-Zr and Zircaloy is dominated by thermal hopping, with negligible contribution from quantum tunneling. The diffusivity predicted by this DFT + KMC approach agrees well with that from previous independent experiments and theories, without using any data fitting. The diffusivity along is found to be slightly higher than that along , with the anisotropy saturated at about 1.20 at high temperatures, resolving contradictory results in previous experiments. Demonstrated using hydrogen diffusion in α-Zr, the same method can be extended for on-lattice diffusion in hcp metals, or systems with similar trapping basins.

Anisotropic hydrogen diffusion in α-Zr and Zircaloy predicted by accelerated kinetic Monte Carlo simulations

PubMed Central

Zhang, Yongfeng; Jiang, Chao; Bai, Xianming

2017-01-01

This report presents an accelerated kinetic Monte Carlo (KMC) method to compute the diffusivity of hydrogen in hcp metals and alloys, considering both thermally activated hopping and quantum tunneling. The acceleration is achieved by replacing regular KMC jumps in trapping energy basins formed by neighboring tetrahedral interstitial sites, with analytical solutions for basin exiting time and probability. Parameterized by density functional theory (DFT) calculations, the accelerated KMC method is shown to be capable of efficiently calculating hydrogen diffusivity in α-Zr and Zircaloy, without altering the kinetics of long-range diffusion. Above room temperature, hydrogen diffusion in α-Zr and Zircaloy is dominated by thermal hopping, with negligible contribution from quantum tunneling. The diffusivity predicted by this DFT + KMC approach agrees well with that from previous independent experiments and theories, without using any data fitting. The diffusivity along is found to be slightly higher than that along , with the anisotropy saturated at about 1.20 at high temperatures, resolving contradictory results in previous experiments. Demonstrated using hydrogen diffusion in α-Zr, the same method can be extended for on-lattice diffusion in hcp metals, or systems with similar trapping basins. PMID:28106154

Anisotropic hydrogen diffusion in α-Zr and Zircaloy predicted by accelerated kinetic Monte Carlo simulations

DOE PAGES

Zhang, Yongfeng; Jiang, Chao; Bai, Xianming

2017-01-20

Here, this report presents an accelerated kinetic Monte Carlo (KMC) method to compute the diffusivity of hydrogen in hcp metals and alloys, considering both thermally activated hopping and quantum tunneling. The acceleration is achieved by replacing regular KMC jumps in trapping energy basins formed by neighboring tetrahedral interstitial sites, with analytical solutions for basin exiting time and probability. Parameterized by density functional theory (DFT) calculations, the accelerated KMC method is shown to be capable of efficiently calculating hydrogen diffusivity in α-Zr and Zircaloy, without altering the kinetics of long-range diffusion. Above room temperature, hydrogen diffusion in α-Zr and Zircaloy ismore » dominated by thermal hopping, with negligible contribution from quantum tunneling. The diffusivity predicted by this DFT + KMC approach agrees well with that from previous independent experiments and theories, without using any data fitting. The diffusivity along < c > is found to be slightly higher than that along < a >, with the anisotropy saturated at about 1.20 at high temperatures, resolving contradictory results in previous experiments. Demonstrated using hydrogen diffusion in α-Zr, the same method can be extended for on-lattice diffusion in hcp metals, or systems with similar trapping basins.« less
a Study on the Fretting Fatigue Life of Zircaloy Alloys

NASA Astrophysics Data System (ADS)

Kwon, Jae-Do; Park, Dae-Kyu; Woo, Seung-Wan; Chai, Young-Suck

Studies on the strength and fatigue life of machines and structures have been conducted in accordance with the development of modern industries. In particular, fine and repetitive cyclic damage occurring in contact regions has been known to have an impact on fretting fatigue fractures. The main component of zircaloy alloy is Zr, and it possesses good mechanical characteristics at high temperatures. This alloy is used in the fuel rod material of nuclear power plants because of its excellent resistance. In this paper, the effect of the fretting damage on the fatigue behavior of the zircaloy alloy is studied. Further, various types of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests are performed with a flat-flat contact configuration using a bridge-type contact pad and plate-type specimen. Through these experiments, it is found that the fretting fatigue strength decreases by about 80% as compared to the plain fatigue strength. Oblique cracks are observed in the initial stage of the fretting fatigue, in which damaged areas are found. These results can be used as the basic data for the structural integrity evaluation of corrosion-resisting alloys considering the fretting damages.
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
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.
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.
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.
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.
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.
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.
Transition joints between Zircaloy-2 and stainless steel by diffusion bonding

NASA Astrophysics Data System (ADS)

Bhanumurthy, K.; Krishnan, J.; Kale, G. B.; Banerjee, S.

1994-11-01

The diffusion bonding between Zircaloy-2 and stainless steel (AISI 304L) using niobium, nickel and copper as intermediate layers has been investigated in the temperature range of 750 to 900°C. Bonding was carried out in a vacuum hot press, under compressive loading. Electron probe microanalysis and metallographic analysis showed a good metallurgical compatibility and also indicated the absence of discontunities, micropores and intermetallic compounds at various interfaces. The bond strength of the diffusion bonded assembly was found to be about 400 MPa for the couples bonded at 870°C for 2 h. The dimple structure on the fractured surface is indicative of the ductile mode of failure of the bonded assembly.
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.
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
In situ monitored in-pile creep testing of zirconium alloys

NASA Astrophysics Data System (ADS)

Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

2014-01-01

The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.
Etude du comportement et de la modélisation viscoplastique du zircaloy 4recristallisé sous chargements monotones et cycliques uni et multiaxes

NASA Astrophysics Data System (ADS)

Delobelle, P.; Robinet, P.

1994-08-01

The results of experiment performed on a recrystallized zircaloy 4 alloy in the intermediate temperature domain 20 leqslant T leqslant 400 ^{circ}C are presented. To characterize the anisotropy, especially at 350 ^{circ}C, the tests were made under both monotonic and cyclic uni- and bidirectional loadings, i.e. tension-compression, tension-torsion and tension-internal pressure tests. The different anisotropy coefficients and especially R^p = \\varepsilon^p_{θθ} /\\varepsilon^ p _ {{^-_-}{^-_-} } seem to be temperature independent. An important feature of the behavior of this alloy in the neighbourhood of 300 ^{circ}C is attributed to the dislocations-point defects interactions (dynamic strain aging), phenomena often observed in the solid solutions. For the 2D cyclic non proportional loadings it is shown that a weak supplementary hardening appears, which is a function of the degree of the phase lag. We propose to particularize and to apply a unified viscoplastic model with internal variables to the considered alloy, as the model as already been developed and identified elsewhere for other isotropic materials. From a general point of view the introduction of the anisotropy in the model is made by four tensors of rank 4 ; [ M] is assigned to the flow directions, [ N] to the linear parts of the kinematical hardening variables and [ Q] , [ R] respectively to the dynamic and static recoveries of these tensorial variables. This phenomenological formulation leads to a correct representation of the set of the experimental results presented at 350 ^{circ}C, which provides an a posteriori confirmation of the formalism used. On étudie, entre 20 et 400 ^{circ}C, à l'aide d'essais sous chargements multiaxiaux monotones et cycliques (traction, torsion et pression interne) les propriétés viscoplastiques anisotropes de tube de zircaloy 4 recristallisé. A la température de 350 ^{circ}C, l'anisotropie a été quantifiée de façon détaillée. Les quelques résultats obtenus à la température ambiante ainsi que l'indépendance du rapport R^p = \\varepsilon^p_{θθ}/\\varepsilon^ p_{{^-_-}{^-_-} } avec la température laissent supposer que l'ensemble des coefficients d'anisotropie ne dépendent pas de la température. Par contre, la fluidité de cet alliage présente un minimum très marqué au voisinage de 300 ^{circ}C. Ce comportement est imputable au vieillissement dynamique fréquemment observé dans les solutions solides d'insertion. Lors d'un chargement cyclique hors phase (traction-torsion déphasée à 90^{circ}) ce matériau présente un léger durcissement supplémentaire. On propose l'extension au cas du zircaloy 4 de la formulation d'un modèle viscoplastique unifié développé et identifié par ailleurs sur d'autres matériaux initialement isotropes. D'une manière générale, l'introduction de l'anisotropie dans ce modèle s'effectue par l'intermédiaire de quatre tenseurs d'ordre 4 affectant les directions d'écoulement [ M] , les parties linéaires des écrouissages cinématiques [ N] , ainsi que les restaurations dynamiques [ Q] et statiques [ R] de ces mêmes variables d'écrouissage. L'identification de ce modèle est discutée et réalisée à 350 ^{circ}C. On montre l'adéquation du formalisme à appréhender l'ensemble des caractéristiques mécaniques de cet alliage.
Hydrogen pickup mechanism of zirconium alloys

NASA Astrophysics Data System (ADS)

Couet, Adrien

Although the optimization of zirconium based alloys has led to significant improvements in hydrogen pickup and corrosion resistance, the mechanisms by which such alloy improvements occur are still not well understood. In an effort to understand such mechanisms, a systematic study of the alloy effect on hydrogen pickup is conducted, using advanced characterization techniques to rationalize precise measurements of hydrogen pickup. The hydrogen pick-up fraction is accurately measured for a specially designed set of commercial and model alloys to investigate the effects of alloying elements, microstructure and corrosion kinetics on hydrogen uptake. Two different techniques to measure hydrogen concentrations were used: a destructive technique, Vacuum Hot Extraction, and a non-destructive one, Cold Neutron Prompt Gamma Activation Analysis. The results indicate that hydrogen pickup varies not only from alloy to alloy but also during the corrosion process for a given alloy. For instance Zircaloy type alloys show high hydrogen pickup fraction and sub-parabolic oxidation kinetics whereas ZrNb alloys show lower hydrogen pickup fraction and close to parabolic oxidation kinetics. Hypothesis is made that hydrogen pickup result from the need to balance charge during the corrosion reaction, such that the pickup of hydrogen is directly related to (and indivisible of) the corrosion mechanism and decreases when the rate of electron transport or oxide electronic conductivity sigmao xe through the protective oxide increases. According to this hypothesis, alloying elements (either in solid solution or in precipitates) embedded in the oxide as well as space charge variations in the oxide would impact the hydrogen pick-up fraction by modifying sigmaox e, which drives oxidation and hydriding kinetics. Dedicated experiments and modelling were performed to assess and validate these hypotheses. In-situ electrochemical impedance spectroscopy (EIS) experiments were performed on Zircaloy-4 tubes to directly measure the evolution of sigma oxe as function of exposure time. The results show that sigmao xe decreases as function of exposure time and that its variations are directly correlated to the instantaneous hydrogen pickup fraction variations. The electron transport through the oxide layer is thus altered as the oxide grows, reasons for which are yet to be exactly determined. Preliminary results also show that sigma oxe of ZrNb alloys would be much higher compared with Zircaloy-4. Thus, it is confirmed that sigmaox e is a key parameter in the hydrogen and oxidation mechanism. Because the mechanism whereby alloying elements are incorporated into the oxide layer is critical to changing sigmao xe, the evolution of the oxidation state of two common alloying elements, Fe and Nb, when incorporated into the growing oxide layers is investigated using X-Ray Absorption Near-Edge Spectroscopy (XANES) using micro-beam synchrotron radiation on cross sectional oxide samples. The results show that the oxidation of both Fe and Nb is delayed in the oxide layer compared to that of Zr, and that this oxidation delay is related to the variations of the instantaneous hydrogen pick-up fraction with exposure time. The evolution of Nb oxidation as function of oxide depth is also compatible with space charge compensation in the oxide and with an increase in sigmaox e of ZrNb alloys compared to Zircaloys. Finally, various successively complex models from the well-known Wagner oxidation theory to the more complex effect of space charge on oxidation kinetics have been developed. The general purpose of the modeling effort is to provide a rationale for the sub-parabolic oxidation kinetics and demonstrate the correlation with hydrogen pickup fraction. It is directly demonstrated that parabolic oxidation kinetics is associated with high sigmao xe and low space charges in the oxide whereas sub-parabolic oxidation kinetics is associated with lower sigmaox e and higher space charge in the oxide. All these observations helped us to propose a general corrosion mechanism of zirconium alloys involving both oxidation and hydrogen pickup mechanism to better understand and predict the effect of alloying additions on the behavior of zirconium alloys.
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.
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
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
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...
Nuclear fuel elements having a composite cladding

DOEpatents

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

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.
Analysis of 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
Characterisation of metallic glass incorporated Zircaloy-2 weldments

NASA Astrophysics Data System (ADS)

Mishra, S.; Savalia, R. T.; Bhanumurthy, K.; Dey, G. K.; Banerjee, S.

1995-12-01

In this study the effect of incorporation of Zr based Fe and Ni bearing metallic glass in spot welds in Zircaloy components has been examined. A comparison of strength and microstructure of the welded joint with and without glass has been carried out. The welded joint with metallic glass has been found to be stronger than the one without metallic glass. The microstructure of the welded region with metallic glass has been found to comprise a large region having martensite. This large martensitic region has also been found to have considerable amount of excess solute (Fe, Ni). The higher strength of the weld with metallic glass seems to originate due to solid solution strengthening, small grain size and the presence of martensitic structure over a large region.
Crystal plasticity modeling of irradiation growth in Zircaloy-2

DOE PAGES

Patra, Anirban; Tome, Carlos; Golubov, Stanislav I.

2017-05-10

A reaction-diffusion based mean field rate theory model is implemented in the viscoplastic self-consistent (VPSC) crystal plasticity framework to simulate irradiation growth in hcp Zr and its alloys. A novel scheme is proposed to model the evolution (both number density and radius) of irradiation-induced dislocation loops that can be informed directly from experimental data of dislocation density evolution during irradiation. This framework is used to predict the irradiation growth behavior of cold-worked Zircaloy-2 and trends compared to available experimental data. The role of internal stresses in inducing irradiation creep is discussed. Effects of grain size, texture, and external stress onmore » the coupled irradiation growth and creep behavior are also studied.« less
Crystal plasticity modeling of irradiation growth in Zircaloy-2

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

Patra, Anirban; Tome, Carlos; Golubov, Stanislav I.

A reaction-diffusion based mean field rate theory model is implemented in the viscoplastic self-consistent (VPSC) crystal plasticity framework to simulate irradiation growth in hcp Zr and its alloys. A novel scheme is proposed to model the evolution (both number density and radius) of irradiation-induced dislocation loops that can be informed directly from experimental data of dislocation density evolution during irradiation. This framework is used to predict the irradiation growth behavior of cold-worked Zircaloy-2 and trends compared to available experimental data. The role of internal stresses in inducing irradiation creep is discussed. Effects of grain size, texture, and external stress onmore » the coupled irradiation growth and creep behavior are also studied.« 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.
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.
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.
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.
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.
Growth and microstructure formation of isothermally-solidified Zircaloy-4 joints brazed by a Zr-Ti-Cu-Ni amorphous alloy ribbon

NASA Astrophysics Data System (ADS)

Kim, K. H.; Lim, C. H.; Lee, J. G.; Lee, M. K.; Rhee, C. K.

2013-10-01

The microstructure and growth characteristics of Zircaloy-4 joints brazed by a Zr48Ti16Cu17Ni19 (at.%) amorphous filler metal have been investigated with regard to the controlled isothermal solidification and intermetallic formation. Two typical joints were produced depending on the isothermal brazing temperature: (1) a dendritic growth structure including bulky segregation in the central zone (at 850 °C), and (2) a homogeneous dendritic structure throughout the joint without segregation (at 890 °C). The primary α-Zr phase was solidified isothermally, nucleating to grow into a joint with a cellular or dendritic structure. Also, the continuous Zr2Ni and particulate Zr2Cu phases were formed in the segregated center zone and at the intercellular region, respectively, owing to the different solubility and atomic mobility of the solute elements (Ti, Cu, and Ni) in the α-Zr matrix. A disappearance of the central Zr2Ni phase was also rate-controlled by the outward diffusion of the Cu and Ni elements. When the detrimental Zr2Ni intermetallic phase was eliminated by a complete isothermal solidification at 890 °C, the strengths of the joints were high enough to cause yielding and fracture in the base metal, exceeding those of the bulk Zircaloy-4, at room temperature as well as at elevated temperatures (up to 400 °C).
High-temperature steam oxidation and oxide crack effects of Zr-1Nb-1Sn-0.1Fe fuel cladding

NASA Astrophysics Data System (ADS)

Lee, Cheol Min; Mok, Yong-Kyoon; Sohn, Dong-Seong

2017-12-01

In this study, high-temperature steam oxidation experiments were performed at 1012-1207 °C on Zr-1Nb-1Sn-0.1Fe fuel cladding tubes to study their weight gains and microstructural characteristics. Many specimens were tested at each test temperature, and the results were reproducible and reliable. It is often debated whether the Zr-1Nb-1Sn-0.1Fe alloy follows the weight gain correlation developed by Cathcart and Pawel (C-P correlation) at around 1000 °C. According to our results, the C-P correlation overpredicts the weight gain at around 1000 °C, and this observation agrees well with the data reported by Westinghouse. In addition, the microstructures of the specimens were analyzed using scanning electron microscopy, and it was found that circumferential cracks are formed at the oxide-metal interface only at around 1000 °C. In previous studies, it has been postulated that cracks in the oxide promote the oxidation process, but it appears that the circumferential cracks at the oxide-metal interface decrease the oxidation rate before the breakaway oxidation occurs by disturbing the diffusion of oxygen. The oxidation rate reduction due to the circumferential cracks appears to be the reason for the overprediction of the C-P correlation at around 1000 °C.
Effect of hydrogenation conditions on the microstructure and mechanical properties of zirconium hydride

NASA Astrophysics Data System (ADS)

Muta, Hiroaki; Nishikane, Ryoji; Ando, Yusuke; Matsunaga, Junji; Sakamoto, Kan; Harjo, Stefanus; Kawasaki, Takuro; Ohishi, Yuji; Kurosaki, Ken; Yamanaka, Shinsuke

2018-03-01

Precipitation of brittle zirconium hydrides deteriorate the fracture toughness of the fuel cladding tubes of light water reactor. Although the hydride embrittlement has been studied extensively, little is known about physical properties of the hydride due to the experimental difficulties. In the present study, to elucidate relationship between mechanical properties and microstructure, two δ-phase zirconium hydrides and one ε-phase zirconium hydride were carefully fabricated considering volume changes at the metal-to-hydride transformation. The δ-hydride that was fabricated from α-zirconium exhibits numerous inner cracks due to the large volume change. Analyses of the neutron diffraction pattern and electron backscatter diffraction (EBSD) data show that the sample displays significant stacking faults in the {111} plane and in the pseudo-layered microstructure. On the other hand, the δ-hydride sample fabricated from β-zirconium at a higher temperature displays equiaxed grains and no cracks. The strong crystal orientation dependence of mechanical properties were confirmed by indentation test and EBSD observation. The δ-hydride hydrogenated from α-zirconium displays a lower Young's modulus than that prepared from β-zirconium. The difference is attributed to stacking faults within the {111} plane, for which the Young's modulus exhibits the highest value in the perpendicular direction. The strong influence of the crystal orientation and dislocation density on the mechanical properties should be considered when evaluating hydride precipitates in nuclear fuel cladding.
Steam Oxidation Testing in the Severe Accident Test Station

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

Pint, Bruce A.

After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative wouldmore » significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.« less
2nd Gen FeCrAl ODS Alloy Development For Accident-Tolerant Fuel Cladding

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

Dryepondt, Sebastien N.; Massey, Caleb P.; Edmondson, Philip D.

Extensive research at ORNL aims at developing advanced low-Cr high strength FeCrAl alloys for accident tolerant fuel cladding. One task focuses on the fabrication of new low Cr oxide dispersion strengthened (ODS) FeCrAl alloys. The first Fe-12Cr-5Al+Y 2O 3 (+ ZrO 2 or TiO 2) ODS alloys exhibited excellent tensile strength up to 800 C and good oxidation resistance in steam up to 1400 C, but very limited plastic deformation at temperature ranging from room to 800 C. To improve alloy ductility, several fabrication parameters were considered. New Fe-10-12Cr-6Al gas-atomized powders containing 0.15 to 0.5wt% Zr were procured and ballmore » milled for 10h, 20h or 40h with Y2O3. The resulting powder was then extruded at temperature ranging from 900 to 1050 C. Decreasing the ball milling time or increasing the extrusion temperature changed the alloy grain size leading to lower strength but enhanced ductility. Small variations of the Cr, Zr, O and N content did not seem to significantly impact the alloy tensile properties, and, overall, the 2nd gen ODS FeCrAl alloys showed significantly better ductility than the 1st gen alloys. Tube fabrication needed for fuel cladding will require cold or warm working associated with softening heat treatments, work was therefore initiated to assess the effect of these fabrications steps on the alloy microstructure and properties. This report has been submitted as fulfillment of milestone M3FT 16OR020202091 titled, Report on 2nd Gen FeCrAl ODS Alloy Development for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.« less
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.
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.
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.
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
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.
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
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.
Performance of U3Si2 Fuel in a Reactivity Insertion Accident

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

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less
A study into the impact of interface roughness development on mechanical degradation of oxides formed on zirconium alloys

NASA Astrophysics Data System (ADS)

Platt, P.; Wedge, S.; Frankel, P.; Gass, M.; Howells, R.; Preuss, M.

2015-04-01

As a cladding material used to encapsulate nuclear fuel pellets, zirconium alloys are the primary barrier separating the fuel and a pressurised steam or lithiated water environment. Degradation mechanisms such as oxidation can be the limiting factor in the life-time of the fuel assembly. Key to controlling oxidation, and therefore allowing increased burn-up of fuel, is the development of a mechanistic understanding of the corrosion process. In an autoclave, the oxidation kinetics for zirconium alloys are typically cyclical, with periods of accelerated kinetics being observed in steps of ∼2 μm oxide growth. These periods of accelerated oxidation are immediately preceded by the development of a layer of lateral cracks near the metal-oxide interface, which may be associated with the development of interface roughness. The present work uses scanning electron microscopy to carry out a statistical analysis of changes in the metal-oxide interface roughness between three different alloys at different stages of autoclave oxidation. The first two alloys are Zircaloy-4 and ZIRLO™ for which analysis is carried out at stages before, during and after first transition. The third alloy is an experimental low tin alloy, which under the same oxidation conditions and during the same time period does not appear to go through transition. Assessment of the metal-oxide interface roughness is primarily carried out based on the root mean square of the interface slope known as the Rdq parameter. Results show clear trends with relation to transition points in the corrosion kinetics. Discussion is given to how this relates to the existing mechanistic understanding of the corrosion process, and the components required for possible future modelling approaches.
Microstructural Characterization of Irradiated U0.7ZrH1.6 Using Ultrasonic Techniques

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

Ramuhalli, Pradeep; Jacob, Richard E.; MacFarlan, Paul J.

In recent years, there has been an increased level of effort to understand the changes in microstructure that occur due to irradiation of nuclear fuel. The primary driver for this increased effort is the potential for designing new fuels that are safer and more reliable, in turn enabling new and improved reactor technologies. Much of the data on microstructural change in irradiated fuels is generated through a host of post irradiation examination techniques such as optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM) to determine grain structure, porosity, crack geometry, etc. in irradiated fuels. Such “traditional”more » examination techniques were recently used to characterize a novel new fuel consisting of U0.17ZrH1.6 pellets bonded to zircaloy-2 cladded with lead-bismuth eutectic before and after irradiation. However, alternative methods such as ultrasonic inspection can provide an opportunity for nondestructively assessing microstructure in both in-pile and post-irradiation examinations. In this paper, we briefly describe initial results of ultrasonic examination of the U0.17ZrH1.6 pellets (unirradiated and irradiated), in a post-irradiation examination study. Data indicate some correlation with microstructural changes due to irradiation; however, it is not clear what the specific microstructural changes are that are influencing the ultrasonic measurements. Interestingly, specimens with nominally identical burnup show differences in ultrasonic signatures, indicating apparent microstructural differences between these specimens. A summary of the experimental study, preliminary data and findings are presented in this short paper. Additional details of the analysis will be included in the presentation.« less
Alternate Fuel Cycle Technologies/Thorium Fuel Cycle Technology Programs. Quarterly report for period 1 April--30 June 1978

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

Vondra, B.L.

1978-08-01

Voloxidation and dissolution studies: rotary-kiln heat-transfer tests are under way using a small rotary kiln along with the development of a mathematical model to determine kiln-heat-flux profiles necessary to maintain a desired temperature gradient. The erosion/corrosion test for evaluating materials of construction is operational. Fuel from a BWR (Big Rock Point) yielded more fine solid residue on dissolution than in previous tests with PWR fuel. Two additional parametric voloxidation tests with H.B. Robinson fuel compared air vs pure oxygen atmospheres at 550{sup 0}C; overall tritium release and subsequent fuel dissolution were equivalent. Thorium dissolution studies: the dissolution rate of thoriamore » in fluoride-catalyzed 8 to 14 M HNO{sub 3} (100{sup 0}C) was max between 0.04 to 0.06 M HF; at higher fluoride concentrations, ThF{sub 4}.5H{sub 2}O precipitated. The rate of zircaloy dissolution continued to increase with increasing fluoride concentration. Stainless-steel-clad (Th,U)0{sub 2} fuel rods irradiated in the NRX reactor were sheared, voloxidized, and dissolved. {le}10% of the tritium was released during voloxidation in air at 600{sup 0}C. Carbon-14 removal from off-gas and fixation: carbon dioxide removal with Linde 13X molecular sieves to less than 100 ppB was experimentally verified using 300 ppM CO in air. Decontamination factors from 3000 to 7500 were obtained for CO{sub 2} removal in the gas-slurry stirred-tank reactor with CA(OH){sub 2}.or Ba(0H){sub 2}/sup .8H2O./. With Ba(OH){sub 2}.H{sub 2}0{sup 2} in a fixed-bed column, decontamination factors of about 30,000 were obtained.« less
Ultra-High Temperature Steam Corrosion of Complex Silicates for Nuclear Applications: A Computational Study

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

Rashkeev, Sergey N.; Glazoff, Michael V.; Tokuhiro, Akira

2014-01-01

Stability of materials under extreme conditions is an important issue for safety of nuclear reactors. Presently, silicon carbide (SiC) is being studied as a cladding material candidate for fuel rods in boiling-water and pressurized water-cooled reactors (BWRs and PWRs) that would substitute or modify traditional zircaloy materials. The rate of corrosion of the SiC ceramics in hot vapor environment (up to 2200 degrees C) simulating emergency conditions of light water reactor (LWR) depends on many environmental factors such as pressure, temperature, viscosity, and surface quality. Using the paralinear oxidation theory developed for ceramics in the combustion reactor environment, we estimatedmore » the corrosion rate of SiC ceramics under the conditions representing a significant power excursion in a LWR. It was established that a significant time – at least 100 h – is required for a typical SiC braiding to significantly degrade even in the most aggressive vapor environment (with temperatures up to 2200 °C) which is possible in a LWR at emergency condition. This provides evidence in favor of using the SiC coatings/braidings for additional protection of nuclear reactor rods against off-normal material degradation during power excursions or LOCA incidents. Additionally, we discuss possibilities of using other silica based ceramics in order to find materials with even higher corrosion resistance than SiC. In particular, we found that zircon (ZrSiO4) is also a very promising material for nuclear applications. Thermodynamic and first-principles atomic-scale calculations provide evidence of zircon thermodynamic stability in aggressive environments at least up to 1535 degrees C.« less
Finite-element model to predict roll-separation force and defects during rolling of U-10Mo alloys

NASA Astrophysics Data System (ADS)

Soulami, Ayoub; Burkes, Douglas E.; Joshi, Vineet V.; Lavender, Curt A.; Paxton, Dean

2017-10-01

A major goal of the Convert Program of the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) is to enable high-performance research reactors to operate with low-enriched uranium rather than the high-enriched uranium currently used. To this end, uranium alloyed with 10 wt% molybdenum (U-10Mo) represents an ideal candidate because of its stable gamma phase, low neutron caption cross section, acceptable swelling response, and predictable irradiation behavior. However, because of the complexities of the fuel design and the need for rolled monolithic U-10Mo foils, new developments in processing and fabrication are necessary. This study used a finite-element code, LS-DYNA, as a predictive tool to optimize the rolling process. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel were conducted following two different schedules. Model predictions of the roll-separation force and roll pack thicknesses at different stages of the rolling process were compared with experimental measurements. The study reported here discussed various attributes of the rolled coupons revealed by the model (e.g., waviness and thickness non-uniformity like dog-boning). To investigate the influence of the cladding material on these rolling defects, other cases were simulated: hot rolling with alternative can materials, namely, 304 stainless steel and Zircaloy-2, and bare-rolling. Simulation results demonstrated that reducing the mismatch in strength between the coupon and can material improves the quality of the rolled sheet. Bare-rolling simulation results showed a defect-free rolled coupon. The finite-element model developed and presented in this study can be used to conduct parametric studies of several process parameters (e.g., rolling speed, roll diameter, can material, and reduction).
Ten inch Planar Optic Display

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

Beiser, L.; Veligdan, J.

A Planar Optic Display (POD) is being built and tested for suitability as a high brightness replacement for the cathode ray tube, (CRT). The POD display technology utilizes a laminated optical waveguide structure which allows a projection type of display to be constructed in a thin (I to 2 inch) housing. Inherent in the optical waveguide is a black cladding matrix which gives the display a black appearance leading to very high contrast. A Digital Micromirror Device, (DMD) from Texas Instruments is used to create video images in conjunction with a 100 milliwatt green solid state laser. An anamorphic opticalmore » system is used to inject light into the POD to form a stigmatic image. In addition to the design of the POD screen, we discuss: image formation, image projection, and optical design constraints.« less
Low-loss hollow-core silica fibers with adjacent nested anti-resonant tubes.

PubMed

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

2015-06-29

We report on numerical design optimization of hollow-core anti-resonant fibers with the aim of reducing transmission losses. We show that re-arranging the nested anti-resonant tubes in the cladding to be adjacent has the effect of significantly reducing leakage as well as bending losses, and for reaching high loss extinction ratios between the fundamental mode and higher order modes. We investigate two versions of the proposed design, one optimized for the mid-IR and another scaled down version for the near-IR and compare them in detail with previously proposed anti-resonant fiber designs including nested elements. Our proposed design is superior with respect to obtaining the lowest leakage losses and the bend losses are also much lower than for the previous designs. Leakage losses as low as 0.0015 dB/km and bending losses of 0.006 dB/km at 5 cm bending radius are predicted at the ytterbium lasing wavelength 1.06 µm. When optimizing the higher-order-mode extinction ratio, the low leakage loss is sacrificed to get an effective single-mode behavior of the fiber. We show that the higher-order-mode extinction ratio is more than 1500 in the range 1.0-1.1 µm around the ytterbium lasing wavelength, while in the mid-IR it can be over 100 around λ = 2.94 μm. This is higher than the previously considered structures in the literature using nested tubes.
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
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
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.
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.
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.
Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

NASA Astrophysics Data System (ADS)

Ohishi, Yuji; Kondo, Toshiki; Ishikawa, Takehiko; Okada, Junpei T.; Watanabe, Yuki; Muta, Hiroaki; Kurosaki, Ken; Yamanaka, Shinsuke

2017-03-01

It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr1-xNix (x = 0.12 and 0.24) and Zr0.77Cr0.23) using the electrostatic levitation technique.
AN ATTEMPT TO LOCATE INTERMETALLIC PARTICLES IN ZIRCONIUM ALLOYS USING A BITTER FIGURE TECHNIQUE

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

Cox, B.; Harder, B.R.

1961-10-01

The compound ZrFe/sub 2/ is known to be ferromagnetic, and an attempt to locate particles of magnetic material in zircaloy-2 and dilute Zr- Fe alloys by a Bitter figure technlque is described. An Fe/sub 3/O/sub 4/ sol in water-soluble plastic was used to prepare Bitter figures of the alloy surfaces in the form of replicas, which were then examined in an electron microscope. No magnetic particles were located in either zircaloy-2 or a Zr-O.3% Fe alloy. Subsequent work on specimens of ZrFe/sub 2/ showed that the failure to detect it in the dilute alloys arose because the size of themore » intermetallic particles in the latter was smaller than the size of the magnetic domains. (auth)« less
In situ high temperature oxidation analysis of Zircaloy-4 using acoustic emission coupled with thermogravimetry

NASA Astrophysics Data System (ADS)

Omar, Al Haj; Véronique, Peres; Eric, Serris; François, Grosjean; Jean, Kittel; François, Ropital; Michel, Cournil

2015-06-01

Zircaloy-4 oxidation behavior at high temperature (900 °C), which can be reached in case of severe accidental situations in nuclear pressurised water reactor, was studied using acoustic emission analysis coupled with thermogravimetry. Two different atmospheres were used to study the oxidation of Zircaloy-4: (a) helium and pure oxygen, (b) helium and oxygen combined with slight addition of air. The experiments with 20% of oxygen confirm the dependence on oxygen anions diffusion in the oxide scale. Under a mixture of oxygen and air in helium, an acceleration of the corrosion was observed due to the detrimental effect of nitrogen. The kinetic rate increased significantly after a kinetic transition (breakaway). This acceleration was accompanied by an acoustic emission activity. Most of the acoustic emission bursts were recorded after the kinetic transition (post-transition) or during the cooling of the sample. The characteristic features of the acoustic emission signals appear to be correlated with the different populations of cracks and their occurrence in the ZrO2 layer or in the α-Zr(O) layer. Acoustic events were recorded during the isothermal dwell time at high temperature under air. They were associated with large cracks in the zirconia porous layer. Acoustic events were also recorded during cooling after oxidation tests both under air or oxygen. For the latter, cracks were observed in the oxygen enriched zirconium metal phase and not in the dense zirconia layer after 5 h of oxidation.
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.
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.
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
Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

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

Marshall, Margaret A.; Bess, John D.

2015-02-01

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

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

Bretscher, M. M.

1998-10-14

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less
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
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.
Ultra-high sensitivity Fabry-Perot interferometer gas refractive index fiber sensor based on photonic crystal fiber and Vernier effect.

PubMed

Quan, Mingran; Tian, Jiajun; Yao, Yong

2015-11-01

An ultra-high sensitivity open-cavity Fabry-Perot interferometer (FPI) gas refractive index (RI) sensor based on the photonic crystal fiber (PCF) and Vernier effect is proposed and demonstrated. The sensor is prepared by splicing a section of PCF to a section of fiber tube fused with a section of single mode fiber. The air holes running along the cladding of the PCF enable the gas to enter or leave the cavity freely. The reflection beam from the last end face of the PCF is used to generate the Vernier effect, which significantly improves the sensitivity of the sensor. Experimental results show that the proposed sensor can provide an ultra-high RI sensitivity of 30899 nm/RIU. This sensor has potential applications in fields such as gas concentration analyzing and humidity monitoring.
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.
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.
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
In situ synchrotron X-ray diffraction study of hydrides in Zircaloy-4 during thermomechanical cycling

DOE PAGES

Cinbiz, Mahmut N.; Koss, Donald A.; Motta, Arthur T.; ...

2017-02-20

The d-spacing evolution of both in-plane and out-of-plane hydrides has been studied using in situ synchrotron radiation X-ray diffraction during thermo-mechanical cycling of cold-worked stress-relieved Zircaloy-4. The structure of the hydride precipitates is such that the δ{111} d-spacing of the planes aligned with the hydride platelet face is greater than the d-spacing of the 111 planes aligned with the platelet edges. Upon heating from room temperature, the δ{111} planes aligned with hydride plate edges exhibit bi-linear thermally-induced expansion. In contrast, the d-spacing of the (111) plane aligned with the hydride plate face initially contracts upon heating. Furthermore, these experimental resultsmore » can be understood in terms of a reversal of stress state associated with precipitating or dissolving hydride platelets within the α-zirconium matrix.« less
Intergrannular strain evolution in a zircaloy-4 alloy with Widmanstatten microstructure

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

Clausen, Bjorn; Vogel, Sven C; Garlea, Eena

2009-01-01

A Zircaloy-4 alloy with Widmanstatten-Basketweave microstructure and random texture has been used to study the deformation systems responsible for the polycrystalline plasticity at the grain level. The evolution of internal strain and bulk texture is investigated using neutron diffraction and an elasto-plastic self-consistent (EPSC) modeling scheme. The macroscopic stress-strain behavior and intergranular (hkil-specific) strain development, parallel and perpendicular to the loading direction, were measured in-situ during uniaxial tensile loading. Then, the EPSC model was employed to simulate the experimental results. This modeling scheme accounts for the thermal anisotropy; elastic-plastic properties of the constituent grains; and activation, reorientation, and stress relaxationmore » associated with twinning. The agreement between the experiment and the model will be discussed as well as the critical resolved shear stresses (CRSS) and the hardening coefficients obtained from the model.« less
Study of the mechanical behavior of the hydride blister/rim structure in Zircaloy-4 using in-situ synchrotron X-ray diffraction

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

Lin, Jun-li; Han, Xiaochun; Heuser, Brent J.

2016-04-01

High-energy synchrotron X-ray diffraction was utilized to study the mechanical response of the f.c.c delta hydride phase, the intermetallic precipitation with hexagonal C14 lave phase and the alpha-Zr phase in the Zircaloy-4 materials with a hydride rim/blister structure near one surface of the material during in-situ uniaxial tension experiment at 200 degrees C. The f.c.c delta was the only hydride phase observed in the rim/blister structure. The conventional Rietveld refinement was applied to measure the macro-strain equivalent response of the three phases. Two regions were delineated in the applied load versus lattice strain measurement: a linear elastic strain region andmore » region that exhibited load partitioning. Load partitioning was quantified by von Mises analysis. The three phases were observed to have similar elastic modulus at 200 degrees C.« less
Analysis of hardening behavior of sheet metals by a new simple shear test method taking into account the Bauschinger effect

NASA Astrophysics Data System (ADS)

Bang, Sungsik; Rickhey, Felix; Kim, Minsoo; Lee, Hyungyil; Kim, Naksoo

2013-12-01

In this study we establish a process to predict hardening behavior considering the Bauschinger effect for zircaloy-4 sheets. When a metal is compressed after tension in forming, the yield strength decreases. For this reason, the Bauschinger effect should be considered in FE simulations of spring-back. We suggested a suitable specimen size and a method for determining the optimum tightening torque for simple shear tests. Shear stress-strain curves are obtained for five materials. We developed a method to convert the shear load-displacement curve to the effective stress-strain curve with FEA. We simulated the simple shear forward/reverse test using the combined isotropic/kinematic hardening model. We also investigated the change of the load-displacement curve by varying the hardening coefficients. We determined the hardening coefficients so that they follow the hardening behavior of zircaloy-4 in experiments.
Ratcheting fatigue behavior of Zircaloy-2 at room temperature

NASA Astrophysics Data System (ADS)

Rajpurohit, R. S.; Sudhakar Rao, G.; Chattopadhyay, K.; Santhi Srinivas, N. C.; Singh, Vakil

2016-08-01

Nuclear core components of zirconium alloys experience asymmetric stress or strain cycling during service which leads to plastic strain accumulation and drastic reduction in fatigue life as well as dimensional instability of the component. Variables like loading rate, mean stress, and stress amplitude affect the influence of asymmetric loading. In the present investigation asymmetric stress controlled fatigue tests were conducted with mean stress from 80 to 150 MPa, stress amplitude from 270 to 340 MPa and stress rate from 30 to 750 MPa/s to study the process of plastic strain accumulation and its effect on fatigue life of Zircaloy-2 at room temperature. It was observed that with increase in mean stress and stress amplitude accumulation of ratcheting strain was increased and fatigue life was reduced. However, increase in stress rate led to improvement in fatigue life due to less accumulation of ratcheting strain.
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
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.
Boron-copper neutron absorbing material and method of preparation

DOEpatents

Wiencek, Thomas C.; Domagala, Robert F.; Thresh, Henry

1991-01-01

A composite, copper clad neutron absorbing material is comprised of copper powder and boron powder enriched with boron 10. The boron 10 content can reach over 30 percent by volume, permitting a very high level of neutron absorption. The copper clad product is also capable of being reduced to a thickness of 0.05 to 0.06 inches and curved to a radius of 2 to 3 inches, and can resist temperatures of 900.degree. C. A method of preparing the material includes the steps of compacting a boron-copper powder mixture and placing it in a copper cladding, restraining the clad assembly in a steel frame while it is hot rolled at 900.degree. C. with cross rolling, and removing the steel frame and further rolling the clad assembly at 650.degree. C. An additional sheet of copper can be soldered onto the clad assembly so that the finished sheet can be cold formed into curved shapes.
Compact cladding-pumped planar waveguide amplifier and fabrication method

DOEpatents

Bayramian, Andy J.; Beach, Raymond J.; Honea, Eric; Murray, James E.; Payne, Stephen A.

2003-10-28

A low-cost, high performance cladding-pumped planar waveguide amplifier and fabrication method, for deployment in metro and access networks. The waveguide amplifier has a compact monolithic slab architecture preferably formed by first sandwich bonding an erbium-doped core glass slab between two cladding glass slabs to form a multi-layer planar construction, and then slicing the construction into multiple unit constructions. Using lithographic techniques, a silver stripe is deposited and formed at a top or bottom surface of each unit construction and over a cross section of the bonds. By heating the unit construction in an oven and applying an electric field, the silver stripe is then ion diffused to increase the refractive indices of the core and cladding regions, with the diffusion region of the core forming a single mode waveguide, and the silver diffusion cladding region forming a second larger waveguide amenable to cladding pumping with broad area diodes.
Characteristics of Ni-Cr-Fe laser clad layers on EA4T steel

NASA Astrophysics Data System (ADS)

Chen, Wenjing; Chen, Hui; Wang, Yongjing; Li, Congchen; Wang, Xiaoli

2017-07-01

The Ni-Cr-Fe metal powder was deposited on EA4T steel by laser cladding technology. The microstructure and chemical composition of the cladding layer were analyzed by optical microscopy (OM), scanning electron microscopy (SEM) and X-ray diffraction (XRD). The bonding ability between the cladding layer and the matrix was measured. The results showed that the bonding between the cladding layer and the EA4T steel was metallurgical bonding. The microstructure of cladding layer was composed of planar crystals, columnar crystals and dendrite, which consisted of Cr2Ni3, γ phase, M23C6 and Ni3B phases. When the powder feeding speed reached 4 g/min, the upper bainite occurred in the heat affected zone (HAZ). Moreover, the tensile strength of the joint increased, while the yield strength and the ductility decreased.
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.
Effect of CeO2 on TiC Morphology in Ni-Based Composite Coating

NASA Astrophysics Data System (ADS)

Cai, Yangchuan; Luo, Zhen; Chen, Yao

2018-03-01

The TiC/Ni composite coating with different content of CeO2 was fabricated on the Cr12MoV steel by laser cladding. The microstructure of cladding layers with the different content of CeO2 from the bottom to the surface is columnar crystal, cellular crystal, and equiaxed crystal. When the content of CeO2 is 0 %, the cladding layer has a coarse and nonuniform microstructure and TiC particles gathering in the cladding layer, and then the wear resistance was reduced. Appropriate rare-earth elements refined and homogenised the microstructure and enhanced the content of carbides, precipitated TiC particles and original TiC particles were spheroidised and refined, the wear resistance of the cladding layer was improved significantly. Excessive rare-earth elements polluted the grain boundaries and made the excessive burning loss of TiC particles that reduced the wear resistance of the cladding layer.
Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

2015-09-01

Low-enrichment (235U < 20 pct) 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 consisted 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 were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then 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. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.
Explosion Clad for Upstream Oil and Gas Equipment

NASA Astrophysics Data System (ADS)

Banker, John G.; Massarello, Jack; Pauly, Stephane

2011-01-01

Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO2 and/or H2S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.
78 FR 9676 - Clad Steel Plate From Japan: Continuation of Antidumping Duty Order

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-11

... hot-rolling of the cladding metal to ensure efficient welding to the basic metal; any other method of... welding (e.g., electrocladding), in which the cladding metal (nickel, chromium, etc.) is applied to the...

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
Corrosion Resistance of Laser Clads of Inconel 625 and Metco 41C

NASA Astrophysics Data System (ADS)

Němeček, Stanislav; Fidler, Lukáš; Fišerová, Pavla

The present paper explores the impact of laser cladding parameters on the corrosion behaviour of the resulting surface. Powders of Inconel 625 and austenitic Metco 41C steel were deposited on steel substrate. It was confirmed that the level of dilution has profound impact on the corrosion resistance and that dilution has to be minimized. However, the chemical composition of the cladding is altered even in the course of the cladding process, a fact which is related to the increase in the substrate temperature. The cladding process was optimized to achieve maximum corrosion resistance. The results were verified and validated using microscopic observation, chemical analysis and corrosion testing.
Development and study the performance of PBA cladding modified fiber optic intrinsic biosensor for urea detection

NASA Astrophysics Data System (ADS)

Botewad, S. N.; Pahurkar, V. G.; Muley, G. G.

2016-05-01

The fabrication and study of a cladding modified fiber optic intrinsic urea biosensor based on evanescent wave absorbance has been presented. The sensor was prepared using cladding modification technique by removing a small portion of cladding of an optical fiber and modifying with an active cladding of porous polyaniline-boric acid (PBA) matrix to immobilize enzyme-urease through cross-linking via glutaraldehyde. The nature of as-synthesized and deposited PBA film on fiber optic sensing element was studied by ultraviolet-visible (UV-vis) spectroscopy and X-ray diffraction (XRD) analysis. The performance of the developed sensor was studied for different urea concentrations in solutions prepared in phosphate buffer.
Effects of Synchronous Rolling on Microstructure, Hardness, and Wear Resistance of Laser Multilayer Cladding

NASA Astrophysics Data System (ADS)

Zhao, W.; Zha, G. C.; Xi, M. Z.; Gao, S. Y.

2018-03-01

A synchronous rolling method was proposed to assist laser multilayer cladding, and the effects of this method on microstructure, microhardness, and wear resistance were studied. Results show that the microstructure and mechanical properties of the traditional cladding layer exhibit periodic inhomogeneity. Synchronous rolling breaks the columnar dendrite crystals to improve the uniformity of the organization, and the residual plastic energy promotes the precipitation of strengthening phases, as CrB, M7C3, etc. The hardness and wear resistance of the extruded cladding layer increase significantly because of the grain refinement, formation of dislocations, and dispersion strengthening. These positive significances of synchronous rolling provide a new direction for laser cladding technology.
High power cladding light stripper using segmented corrosion method: theoretical and experimental studies.

PubMed

Yin, Lu; Yan, Mingjian; Han, Zhigang; Wang, Hailin; Shen, Hua; Zhu, Rihong

2017-04-17

We present the segmented corrosion method that uses hydrofluoric acid to etch the fiber of a fiber laser for removing high-power cladding light to improve stripping uniformity and power handling capability. For theoretical guidelines, we propose a simulation model of etched-fiber stripping to evaluate the relationship between the etched-fiber parameters and cladding light attenuation and to analyze the stripping uniformity achieved with segmented corrosion. A two-segment etched fiber is fabricated with cladding light attenuation of 19.8 dB and power handling capability up to 670 W. We find that the cladding light is stripped uniformly and the temperature distribution is uniform without the formation of hot spots.
Fabrication of versatile cladding light strippers and fiber end-caps with CO2 laser radiation

NASA Astrophysics Data System (ADS)

Steinke, M.; Theeg, T.; Wysmolek, M.; Ottenhues, C.; Pulzer, T.; Neumann, J.; Kracht, D.

2018-02-01

We report on novel fabrication schemes of versatile cladding light strippers and end-caps via CO2 laser radiation. We integrated cladding light strippers in SMA-like connectors for reliable and stable fiber-coupling of high-power laser diodes. Moreover, the application of cladding light strippers in typical fiber geometries for high-power fiber lasers was evaluated. In addition, we also developed processes to fuse end-caps to fiber end faces via CO2 laser radiation and inscribe the fibers with cladding light strippers near the end-cap. Corresponding results indicate the great potential of such devices as a monolithic and low-cost alternative to SMA connectors.
Semipolar III-nitride laser diodes with zinc oxide cladding.

PubMed

Myzaferi, Anisa; Reading, Arthur H; Farrell, Robert M; Cohen, Daniel A; Nakamura, Shuji; DenBaars, Steven P

2017-07-24

Incorporating transparent conducting oxide (TCO) top cladding layers into III-nitride laser diodes (LDs) improves device design by reducing the growth time and temperature of the p-type layers. We investigate using ZnO instead of ITO as the top cladding TCO of a semipolar (202¯1) III-nitride LD. Numerical modeling indicates that replacing ITO with ZnO reduces the internal loss in a TCO clad LD due to the lower optical absorption in ZnO. Lasing was achieved at 453 nm with a threshold current density of 8.6 kA/cm 2 and a threshold voltage of 10.3 V in a semipolar (202¯1) III-nitride LD with ZnO top cladding.
CO₂ laser-fabricated cladding light strippers for high-power fiber lasers and amplifiers.

PubMed

Boyd, Keiron; Simakov, Nikita; Hemming, Alexander; Daniel, Jae; Swain, Robert; Mies, Eric; Rees, Simon; Andrew Clarkson, W; Haub, John

2016-04-10

We present and characterize a simple CO₂ laser processing technique for the fabrication of compact all-glass optical fiber cladding light strippers. We investigate the cladding light loss as a function of radiation angle of incidence and demonstrate devices in a 400 μm diameter fiber with cladding losses of greater than 20 dB for a 7 cm device length. The core losses are also measured giving a loss of <0.008±0.006 dB/cm. Finally we demonstrate the successful cladding light stripping of a 300 W laser diode with minimal heating of the fiber coating and packaging adhesives.
Property Investigation of Laser Cladded, Laser Melted and Electron Beam Melted Ti-Al6-V4

DTIC Science & Technology

2006-05-01

UNCLASSIFIED/UNLIMITED UNCLASSIFIED/UNLIMITED Figure 3: Examples of electron beam melted net shape parts; powder bed [3]. 1.4 Laser Cladding ...description, www.arcam.com. [4] K.-H. Hermann, S. Orban, S. Nowotny, Laser Cladding of Titanium Alloy Ti6242 to Restore Damaged Blades, Proceedings...Property Investigation of Laser Cladded , Laser Melted and Electron Beam Melted Ti-Al6-V4 Johannes Vlcek EADS Deutschland GmbH Corporate Research
Comparison of fiber lasers based on distributed side-coupled cladding-pumped fibers and double-cladding fibers.

PubMed

Huang, Zhihe; Cao, Jianqiu; Guo, Shaofeng; Chen, Jinbao; Xu, Xiaojun

2014-04-01

We compare both analytically and numerically the distributed side-coupled cladding-pumped (DSCCP) fiber lasers and double cladding fiber (DCF) lasers. We show that, through optimization of the coupling and absorbing coefficients, the optical-to-optical efficiency of DSCCP fiber lasers can be made as high as that of DCF lasers. At the same time, DSCCP fiber lasers are better than the DCF lasers in terms of thermal management.
Preliminary calculations related to the accident at Three Mile Island

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

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

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

PubMed Central

Peng, Xian; Yuan, Han; Chen, Wufan; Ding, Lei

2017-01-01

Continuous loop averaging deconvolution (CLAD) is one of the proven methods for recovering transient auditory evoked potentials (AEPs) in rapid stimulation paradigms, which requires an elaborated stimulus sequence design to attenuate impacts from noise in data. The present study aimed to develop a new metric in gauging a CLAD sequence in terms of noise gain factor (NGF), which has been proposed previously but with less effectiveness in the presence of pink (1/f) noise. We derived the new metric by explicitly introducing the 1/f model into the proposed time-continuous sequence. We selected several representative CLAD sequences to test their noise property on typical EEG recordings, as well as on five real CLAD electroencephalogram (EEG) recordings to retrieve the middle latency responses. We also demonstrated the merit of the new metric in generating and quantifying optimized sequences using a classic genetic algorithm. The new metric shows evident improvements in measuring actual noise gains at different frequencies, and better performance than the original NGF in various aspects. The new metric is a generalized NGF measurement that can better quantify the performance of a CLAD sequence, and provide a more efficient mean of generating CLAD sequences via the incorporation with optimization algorithms. The present study can facilitate the specific application of CLAD paradigm with desired sequences in the clinic. PMID:28414803
Parametric Study and Multi-Criteria Optimization in Laser Cladding by a High Power Direct Diode Laser

NASA Astrophysics Data System (ADS)

Farahmand, Parisa; Kovacevic, Radovan

2014-12-01

In laser cladding, the performance of the deposited layers subjected to severe working conditions (e.g., wear and high temperature conditions) depends on the mechanical properties, the metallurgical bond to the substrate, and the percentage of dilution. The clad geometry and mechanical characteristics of the deposited layer are influenced greatly by the type of laser used as a heat source and process parameters used. Nowadays, the quality of fabricated coating by laser cladding and the efficiency of this process has improved thanks to the development of high-power diode lasers, with power up to 10 kW. In this study, the laser cladding by a high power direct diode laser (HPDDL) as a new heat source in laser cladding was investigated in detail. The high alloy tool steel material (AISI H13) as feedstock was deposited on mild steel (ASTM A36) by a HPDDL up to 8kW laser and with new design lateral feeding nozzle. The influences of the main process parameters (laser power, powder flow rate, and scanning speed) on the clad-bead geometry (specifically layer height and depth of the heat affected zone), and clad microhardness were studied. Multiple regression analysis was used to develop the analytical models for desired output properties according to input process parameters. The Analysis of Variance was applied to check the accuracy of the developed models. The response surface methodology (RSM) and desirability function were used for multi-criteria optimization of the cladding process. In order to investigate the effect of process parameters on the molten pool evolution, in-situ monitoring was utilized. Finally, the validation results for optimized process conditions show the predicted results were in a good agreement with measured values. The multi-criteria optimization makes it possible to acquire an efficient process for a combination of clad geometrical and mechanical characteristics control.
Structure and tribological properties of steel after non-vacuum electron beam cladding of Ti, Mo and graphite powders

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

Bataev, I.A.; Mul, D.O.; Bataev, A.A.

2016-02-15

The non-vacuum electron beam cladding technique was used to fabricate layers alloyed with Ti, Mo and C on the surface of low-alloyed steel. Two types of experiments were carried out. In the first experiment, a mixture of Ti and graphite powders was used for cladding; in the second, a mixture of Ti, Mo and graphite powders was used for cladding. CaF{sub 2} powder or a mixture of CaF{sub 2} and LiF powders was used as flux. The thickness of the cladded layers was in the range of 2–2.2 mm. The structure of the layers was studied using optical microscopy, scanningmore » electron microscopy (SEM), transmission electron microscopy (TEM), and X-ray diffraction (XRD). The microhardness after cladding of the layers fabricated by cladding of Ti and graphite powders was 8–9 GPa, while the microhardness of layers with Mo additions reached 11–12 GPa. The highest wear resistance at sliding friction and friction in abrasive environment was reached in the samples fabricated using Ti, Mo and graphite mixture due to the higher hardness and the martensite–austenite structure of the matrix. The wear resistance against fixed abrasive particles was 2.4 times higher compared to that of carburized and quenched steel. - Highlights: • Ti, C and Mo mixture of powders was cladded using non-vacuum electron beam treatment. • The depth of the cladded layers was 2.0 … 2.2 mm. • The microhardness of layer with Mo, Ti and C additions reached ~ 11 … 12 GPa. • The hardening of the layers caused by the formation of TiC particles and martensitic matrix • Wear resistance of cladded coatings was 2.4 higher than carburized steel.« less
Analysis of laser-induction hybrid cladding processing conditions

NASA Astrophysics Data System (ADS)

Huang, Yongjun; Zeng, Xiaoyan; Hu, Qianwu

2007-12-01

A new cladding approach based on laser-induction hybrid technique on flat sheets is presented in this paper. Coating is produced by means of 5kw cw CO II laser equipped with 100kw high frequent inductor, and the experiments set-up, involving a special machining-head, which can provide laser-induction hybrid heat resources simultaneously. The formation of thick NiCrSiB coating on a steel substrate by off-axial powder feeding is studied from an experimental point of view. A substrate melting energy model is developed to describe the energy relationship between laser-induction hybrid cladding and laser cladding alone quantitatively. By comparing the experimental results with the calculational ones, it is shown that the tendency of fusion zone height of theoretical calculation is in agreement with that of tests in laser-induction hybrid cladding. Via analyses and tests, the conclusions can be lead to that the fusion zone height can be increased easily and the good bond of cladding track can be achieved within wide cladding processing window in laser-induction hybrid processing. It shows that the induction heating has an obvious effect on substrate melting and metallurgical bond.
Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

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

Mickalonis, J. I.

2015-08-31

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material withmore » the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.« less
Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

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

Mickalonis, J. I.

2015-08-01

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material withmore » the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.« less
Fuel clad chemical interactions in fast reactor MOX fuels

NASA Astrophysics Data System (ADS)

Viswanathan, R.

2014-01-01

Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel-Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ṡ [B/(at.% fission)] ṡ (T/K-705) ṡ [(O/M)i-1.935]} + 20.5) for (O/M)i ⩽ 1.98. A new model is proposed for (O/M)i ⩾ 1.98: d/μm = [B/(at.% fission)] ṡ (T/K-800)0.5 ṡ [(O/M)i-1.94] ṡ [P/(W cm-1)]0.5. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M)i is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.
77 FR 31834 - Clad Steel Plate from Japan: Final Results of the Expedited Third Sunset Review of the...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-30

... rolling; simple hot-rolling of the cladding metal to ensure efficient welding to the basic metal; any... process to ensure welding (e.g., electrocladding), in which the cladding metal (nickel, chromium, etc.) is...
Diode-Pumped Thulium (Tm)/Holmium (Ho) Composite Fiber 2.1-Micrometers Laser

DTIC Science & Technology

2015-09-01

composite fiber laser of holmium-core and thulium-doped cladding . The composite fiber was optically pumped by an 803-nm fiber coupled diode source and was...4 odd and 5 even modes were exclusive to the core and first cladding . As the Tm laser modes are excluded from lasing in the second (undoped...of the Tm-doped clad /Ho-doped core fiber laser . In particular, calculations of the model overlap of the cladding modes with the core have been

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.
Transversely polarized source cladding for an optical fiber

NASA Technical Reports Server (NTRS)

Egalon, Claudio Oliveira (Inventor); Rogowski, Robert S. (Inventor)

1994-01-01

An optical fiber comprising a fiber core having a longitudinal symmetry axis is provided. An active cladding surrounds a portion of the fiber core and comprises light-producing sources which emit light in response to chemical or light excitation. The cladding sources are oriented transversely with respect to the longitudinal axis of the fiber core. This polarization results in a superior power efficiency compared to active cladding sources that are randomly polarized or longitudinally polarized parallel with the longitudinal symmetry axis.
Effects of the weld thermal cycle on the microstructure of alloy 690

NASA Astrophysics Data System (ADS)

Tuttle, James R.

Alloy 690 has been introduced as a material for use as the heat exchanger tubes in the steam generators (SGs) of pressurised water reactor (PWR) nuclear power plant. Its immediate predecessor, alloy 600, suffered from a number of degradation modes and another alternative, alloy 800, has also had in-service problems. In laboratory tests, alloy 690 in both mill annealed (MA) and special thermally treated (STT) condition has shown a high degree of resistance to degradation in simulated PWR primary side environments and other test media.Limited research has previously been undertaken to investigate the effects of welding on alloy 690, when the material is used in SG applications. It was deemed important to increase knowledge in this area since fabrication of PWR SGs involves gas tungsten arc welding (GTAW) of the heat exchanger tubes to a clad tubeplate. For this research investigation welded samples of alloy 690 have been produced in the laboratory using a range of thermal cycles based around recommended weld parameters for SG fabrication. These samples have been compared with archive welds from PWR SG manufacturers. A number of welds incorporating alloy 600 and a number using alloy 800 tubing material have also been fabricated in the laboratory for comparative purposes. Two experimental melts have been produced to study the effects of Nb substitution for Ti in alloy 690 type materials.Welded and unwelded specimens have been studied, analysed and tested using a variety of methods and techniques. A method of metallographic sample preparation for transmission electron microscope (TEM) thin foil specimens has been developed and documented which ensures foil perforation in a specific region. The effects of Nb substitution for Ti have been discussed. Chemical balances and microstructures in the fusion zone of welds manufactured from alloy 690 tubing incorporating alloy 82 weld consumable have been shown to be non-ideal. Within the heat affected zone (HAZ) of both laboratory produced and archive welds the microstructures have been identified as detrimentally altered from the STT condition original tubing material(s). A number of conclusions have been drawn and recommendations have been made for future work.
Migration of Point Defects in the Field of a Temperature Gradient

NASA Astrophysics Data System (ADS)

Kozlov, A. V.; Portnykh, I. A.; Pastukhov, V. I.

2018-04-01

The influence of the temperature gradient over the thickness of the cladding of a fuel element of a fast-neutron reactor on the migration of point defects formed in the cladding material due to neutron irradiation has been studied. It has been shown that, under the action of the temperature gradient, the flux of vacancies onto the inner surface of the cladding is higher than the flux of interstitial atoms, which leads to the formation of a specific concentration profile in the cladding with a vacancy-depleted zone near the inner surface. The experimental results on the spatial distribution of pores over the cladding thickness have been presented with which the data on the concentration profiles and vacancy fluxes have been compared.
Development and study the performance of PBA cladding modified fiber optic intrinsic biosensor for urea detection

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

Botewad, S. N.; Pahurkar, V. G.; Muley, G. G., E-mail: gajananggm@yahoo.co.in

2016-05-06

The fabrication and study of a cladding modified fiber optic intrinsic urea biosensor based on evanescent wave absorbance has been presented. The sensor was prepared using cladding modification technique by removing a small portion of cladding of an optical fiber and modifying with an active cladding of porous polyaniline-boric acid (PBA) matrix to immobilize enzyme-urease through cross-linking via glutaraldehyde. The nature of as-synthesized and deposited PBA film on fiber optic sensing element was studied by ultraviolet-visible (UV-vis) spectroscopy and X-ray diffraction (XRD) analysis. The performance of the developed sensor was studied for different urea concentrations in solutions prepared in phosphatemore » buffer.« less
Widely tunable asymmetric long-period fiber grating with high sensitivity using optical polymer on laser-ablated cladding.

PubMed

Chen, Nan-Kuang; Hsu, Der-Yi; Chi, Sien

2007-08-01

We demonstrate high-efficiency, wideband-tunable, laser-ablated long-period fiber gratings that use an optical polymer overlay. Portions of the fiber cladding are periodically removed by CO(2) laser pulses to induce periodic index changes for coupling the core mode into cladding modes. An optical polymer with a high thermo-optic coefficient with a dispersion distinct from that of silica is used on a deep-ablated cladding structure so that the effective indices of cladding modes become dispersive and the resonant wavelengths can be efficiently tuned. The tuning efficiency can be as high as 15.8 nm/ degrees C, and the tuning range can be wider than 105 nm (1545-1650 nm).
Integrated double-clad photonic crystal fiber amplifier

NASA Astrophysics Data System (ADS)

Liu, Jun; Gu, Yanran; Chen, Zilun

2017-10-01

This paper studies and fabricates an integrated double-clad photonic crystal fiber amplifier, which overcomes the shortcomings of space application and makes full use of excellent property of double-clad photonic crystal fiber. In the experiment, the (6 + 1) × 1 end-pump coupler with DC-PCF is fabricated. The six pump fibers are fabricated with 105 / 125μm (NA = 0.22) multi-mode fiber. The signal fiber is made of ordinary single-mode fiber SMF-28. Then we spliced the tapered fiber bundle to photonic crystal fiber. At last, we produce double-clad photonic crystal fiber with an end-cap that are able to withstand high average power and protect the system. We have fabricated an integrated Yb-double-clad photonic crystal fiber amplifier.
Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

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

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.
Laser Cladding of TiAl Intermetallic Alloy on Ti6Al4V -Process Optimization and Properties

NASA Astrophysics Data System (ADS)

Cárcel, B.; Serrano, A.; Zambrano, J.; Amigó, V.; Cárcel, A. C.

In order to improve Ti6Al4V high-temperature resistance and its tribological properties, the deposition of TiAl intermetallic (Ti-48Al-2Cr-2Nb) coating on a Ti6Al4V substrate by coaxial laser cladding has been investigated. Laser cladding by powder injection is an emerging laser material processing technique that allows the deposition of thick protective coatings on substrates,using a high power laser beam as heat source. Laser cladding is a multiple-parameter-dependent process. The main process parameters involved (laser power, powder feeding rate, scanning speed and preheating temperature) has been optimized. The microstructure and geometrical quantities (clad area and dilution) of the coating was characterized by optical microscopy and scanning electron microscopy (SEM). In addition the cooling rate of the clad during the process was measured by a dual-color pyrometer. This result has been related to defectology and mechanical coating properties.
Laser Cladding of Ti-6Al-4 V Powder on Ti-6Al-4 V Substrate: Effect of Laser Cladding Parameters on Microstructure

NASA Astrophysics Data System (ADS)

Cottam, Ryan; Brandt, Milan

The laser cladding of Ti-6Al-4 V powder on Ti-6Al-4 V substrate has been investigated to determine laser parameters that could be used as a repair technology for Ti-6Al-4 V components. The parameters chosen for the investigation were developed by an analytical laser cladding model. Holding clad height and melt pool depth constant, the traversing speed was varied between 300 mm/min and 1500 mm/min, an associated power for the given speed was calculated by the model. Two different melt pool depths were used in the calculation of laser power for a given process velocity. The resulting microstructures in the clad zone varied from a relatively thin martensitic structure to a dendritic/thick martensitic structure. The heat affected zone (HAZ) showed a refinement of the Widmanstatten microstructure with a decreasing laser traversing speed and a coarser martensitic structure for the sample prepared with a deeper melt pool.
Microstructure and properties of laser-clad high-temperature wear-resistant alloys

NASA Astrophysics Data System (ADS)

Yang, Yongqiang

1999-02-01

A 2-kW CO 2 laser with a powder feeder was used to produce alloy coatings with high temperature-wear resistance on the surface of steel substrates. To analyze the microstructure and microchemical composition of the laser-clad layers, a scanning electron microscope (SEM) equipped with an energy dispersive X-ray microanalysis system was employed. X-ray diffraction techniques were applied to characterize the phases formed during the cladding process. The results show that the microstructure of the cladding alloy consists mainly of many dispersed particles (W 2C, (W,Ti)C 1- x, WC), a lamellar eutectic carbide M 12C, and an (f.c.c) matrix. Hardness tested at room and high temperature showed that the laser-clad zone has a moderate room temperature hardness and relatively higher elevated temperature hardness. The application of the laser-clad layer to a hot tool was very successful, and its operational life span was prolonged 1 to 4 times.
Investigation of semiconductor clad optical waveguides

NASA Technical Reports Server (NTRS)

Batchman, T. E.; Mcwright, G.

1981-01-01

The properties of semiconductor-clad optical waveguides based on glass substrates were investigated. Computer modeling studies on four-layer silicon-clad planar dielectric waveguides indicated that the attenuation and mode index should behave as exponentially damped sinusoids as the silicon thickness is decreased below one micrometer. This effect can be explained as a periodic coupling between the guided modes of the lossless structure and the lossy modes supported by the high refractive index silicon. The computer studies also show that both the attenuation and mode index of the propagating mode are significantly altered by conductivity charges in the silicon. Silicon claddings were RF sputtered onto AgNO3-NaNO3 ion exchanged waveguides and preliminary measurements of attenuation were made. An expression was developed which predicts the attenuation of the silicon clad waveguide from the attenuation and phase characteristics of a silicon waveguide. Several applications of these clad waveguides are suggested and methods for increasing the photo response of the RF sputtered silicon films are described.
An analytical model to predict and minimize the residual stress of laser cladding process

NASA Astrophysics Data System (ADS)

Tamanna, N.; Crouch, R.; Kabir, I. R.; Naher, S.

2018-02-01

Laser cladding is one of the advanced thermal techniques used to repair or modify the surface properties of high-value components such as tools, military and aerospace parts. Unfortunately, tensile residual stresses generate in the thermally treated area of this process. This work focuses on to investigate the key factors for the formation of tensile residual stress and how to minimize it in the clad when using dissimilar substrate and clad materials. To predict the tensile residual stress, a one-dimensional analytical model has been adopted. Four cladding materials (Al2O3, TiC, TiO2, ZrO2) on the H13 tool steel substrate and a range of preheating temperatures of the substrate, from 300 to 1200 K, have been investigated. Thermal strain and Young's modulus are found to be the key factors of formation of tensile residual stresses. Additionally, it is found that using a preheating temperature of the substrate immediately before laser cladding showed the reduction of residual stress.
Suppression of dilution in Ni-Cr-Si-B alloy cladding layer by controlling diode laser beam profile

NASA Astrophysics Data System (ADS)

Tanigawa, Daichi; Funada, Yoshinori; Abe, Nobuyuki; Tsukamoto, Masahiro; Hayashi, Yoshihiko; Yamazaki, Hiroyuki; Tatsumi, Yoshihiro; Yoneyama, Mikio

2018-02-01

A Ni-Cr-Si-B alloy layer was produced on a type 304 stainless steel plate by laser cladding. In order to produce cladding layer with smooth surface and low dilution, influence of laser beam profile on cladding layer was investigated. A laser beam with a constant spatial intensity at the focus spot was used to suppress droplet formation during the cladding layer formation. This line spot, formed with a focussing unit designed by our group, suppressed droplet generation. The layer formed using this line spot with a constant spatial intensity had a much smoother surface compared to a layer formed using a line spot with a Gaussian-like beam. In addition, the dilution of the former layer was much smaller. These results indicated that a line spot with a constant spatial intensity was more effective in producing a cladding layer with smooth surface and low dilution because it suppressed droplet generation.
Interfacial microstructure and properties of copper clad steel produced using friction stir welding versus gas metal arc welding

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

Shen, Z.; Chen, Y.; Haghshenas, M., E-mail: mhaghshe@uwaterloo.ca

A preliminary study compares the feasibility and microstructures of pure copper claddings produced on a pressure vessel A516 Gr. 70 steel plate, using friction stir welding versus gas metal arc welding. A combination of optical and scanning electron microscopy is used to characterize the grain structures in both the copper cladding and heat affected zone in the steel near the fusion line. The friction stir welding technique produces copper cladding with a grain size of around 25 μm, and no evidence of liquid copper penetration into the steel. The gas metal arc welding of copper cladding exhibits grain sizes overmore » 1 mm, and with surface microcracks as well as penetration of liquid copper up to 50 μm into the steel substrate. Transmission electron microscopy reveals that metallurgical bonding is produced in both processes. Increased diffusion of Mn and Si into the copper cladding occurs when using gas metal arc welding, although some nano-pores were detected in the FSW joint interface. - Highlights: • Cladding of steel with pure copper is possible using either FSW or GMAW. • The FSW yielded a finer grain structure in the copper, with no evidence of cracking. • The FSW joint contains some evidence of nano-pores at the interface of the steel/copper. • Copper cladding by GMAW contained surface cracks attributed to high thermal stresses. • The steel adjacent to the fusion line maintained a hardness value below 248 HV.« less
Metal clad aramid fibers for aerospace wire and cable

NASA Technical Reports Server (NTRS)

Tokarsky, Edward W.; Dunham, Michael G.; Hunt, James E.; Santoleri, E. David; Allen, David B.

1995-01-01

High strength light weight metal clad aramid fibers can provide significant weight savings when used to replace conventional metal wire in aerospace cable. An overview of metal clad aramid fiber materials and information on performance and use in braided electrical shielding and signal conductors is provided.
Investigation of a quadrupole ultra-high vacuum ion pump

NASA Technical Reports Server (NTRS)

Schwarz, H. J.

1974-01-01

The new nonmagnetic ion pump resembles the quadrupole ionization gage. The dimensions are larger, and hyperbolically shaped electrodes replace the four rods. Their surfaces follow y sq. = 36 + x sq. (x, y in centimeters). The electrodes, 55 cm long, are positioned lengthwise in a tube. At one end a cathode emits electrons; at the other end a narrowly wound flat spiral of tungsten clad with titanium on cathode potential can be heated for titanium evaporation. Electrons accelerated by a dc potential of the surface electrodes oscillate between the ends on rotational trajectories, if a high frequency potential superimposed on the dc potential is properly adjusted. Pumping speeds (4-100 liter/sec) for different gases at different peak voltages (1000-3000V) at corresponding frequencies (57-100 MHz), and at different pressures 0.00001 to the minus 9 power Torr were observed. The lowest pressure reached was below 10 to the minus 10 power Torr.
Broadband mid-infrared supercontinuum generation in novel As2Se3-As2Se2 S step-index fibers

NASA Astrophysics Data System (ADS)

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

2018-03-01

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

DOEpatents

Sprehn, G.A.; Hrubesh, L.W.; Poco, J.F.; Sandler, P.H.

1997-11-04

An optical fiber is surrounded by an aerogel cladding. For a low density aerogel, the index of refraction of the aerogel is close to that of air, which provides a high numerical aperture to the optical fiber. Due to the high numerical aperture, the aerogel clad optical fiber has improved light collection efficiency. 4 figs.
Aerogel-clad optical fiber

DOEpatents

Sprehn, Gregory A.; Hrubesh, Lawrence W.; Poco, John F.; Sandler, Pamela H.

1997-01-01

An optical fiber is surrounded by an aerogel cladding. For a low density aerogel, the index of refraction of the aerogel is close to that of air, which provides a high numerical aperture to the optical fiber. Due to the high numerical aperture, the aerogel clad optical fiber has improved light collection efficiency.

Some links on this page may take you to non-federal websites. Their policies may differ from this site.